ITR-18-003

ITER Research Plan within
the Staged Approach(Level III – Provisional Version)
ITER Organization

ITR.support@iter.org

17 September 2018

ITER Research Plan within the Staged Approach(Level III – Provisional Version)

ITR-18-003

ITER Research Plan LIII – Provisional Version
Table of Contents
ACRONYMS.................................................................................................................................................... 8
ACKNOWLEDGEMENTS .......................................................................................................................... 13
EXECUTIVE SUMMARY ........................................................................................................................... 14
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INTRODUCTION ................................................................................................................................... 22

1.1 OBJECTIVES ........................................................................................................................................................... 22
1.2 METHODOLOGY .................................................................................................................................................... 23
1.3 OVERALL STRUCTURE OF THE ITER RESEARCH PLAN PRESENT VERSION ..................................................... 25
1.3.1 ITER RESEARCH AND OPERATIONAL PHASES ............................................................................................................. 25
1.3.2 ASSUMPTIONS ON THE PHASING OF ITER OPERATIONAL CAPABILITIES ................................................................ 27
1.3.3 KEY RESEARCH ISSUES ..................................................................................................................................................... 30
1.3.3.1 Key Research Issues during Operation ............................................................................................................ 30
1.3.3.2 Key Research Issues during Construction ...................................................................................................... 33
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RESEARCH PROGRAM DURING OPERATIONS .......................................................................... 35

2.1 OVERALL STRUCTURE OF THE EXPERIMENTAL PROGRAM DURING OPERATIONS .......................................... 35
2.1.1 INTEGRATION OF THE RESEARCH PLAN INTO THE STAGED APPROACH .................................................................. 36
2.2 PLASMA-WALL INTERACTION (PWI) ISSUES FOR THE RESEARCH PLAN ....................................................... 37
2.2.1 CHARACTERISTICS OF HELIUM OPERATION ................................................................................................................. 37
2.2.2 STEADY-STATE NEAR AND FAR SCRAPE-OFF LAYER (SOL) POWER WIDTHS (1/IP SCALING) ............................ 40
2.2.3 EDGE LOCALIZED MODE (ELM) POWER WIDTH SCALING .......................................................................................... 42
2.2.4 AMMONIA FORMATION DURING N2 SEEDING ................................................................................................................ 43
2.2.5 DIFFERENCES IN MEDIUM-Z IMPURITY SEEDING (N2 VS NE) .................................................................................... 43
2.2.6 TUNGSTEN-RELATED MATERIAL ISSUES ........................................................................................................................ 45
2.2.7 CASTELLATION GAP HEAT LOADS ................................................................................................................................... 46
2.2.8 RADIO-FREQUENCY-ASSISTED WALL CONDITIONING TECHNIQUES .......................................................................... 47
2.2.9 DIVERTOR DETACHMENT CONTROL ............................................................................................................................... 48
2.2.10 PWI/PLASMA BOUNDARY CODE VALIDATION............................................................................................................ 49
2.3 PLASMA DISRUPTION MANAGEMENT ................................................................................................................. 51
2.4 FIRST PLASMA ...................................................................................................................................................... 53
2.4.1 INTEGRATED COMMISSIONING ........................................................................................................................................ 53
2.4.2 FIRST PLASMA.................................................................................................................................................................... 56
2.4.3 ENGINEERING OPERATION............................................................................................................................................... 56
2.5 PRE-FUSION POWER OPERATION PHASE (PFPO) ........................................................................................... 57
2.5.1 OBJECTIVES FOR THE PRE-FUSION POWER OPERATION PHASE ............................................................................... 57
2.5.2 ASSUMPTIONS FOR PRE-FUSION POWER OPERATION ................................................................................................ 60
2.5.3 PLASMA SCENARIOS FOR HYDROGEN AND HELIUM OPERATION ................................................................................ 62
2.5.3.1 First Plasma ................................................................................................................................................................. 64
2.5.3.2 First divertor plasma in PFPO-1 ......................................................................................................................... 64
2.5.3.3 First q95 = 3 plasma in PFPO-1 and PFPO-2 ................................................................................................... 65
2.5.3.4 5 MA/1.8 T scenarios in PFPO-1 and PFPO-2 ............................................................................................... 66
2.5.3.5 Progressive steps towards full current in PFPO-1 and PFPO-2 ............................................................. 67
2.5.3.6 First full current plasma in PFPO-1 and PFPO-2.......................................................................................... 67
2.5.4 OPERATIONS PLAN FOR PFPO-1 .................................................................................................................................... 68

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2.5.4.1 PFPO-1 Objectives .................................................................................................................................................... 68
2.5.4.2 Establishing routine operation with plasma to 3.5 MA divertor configuration .............................. 72
2.5.4.2.1 Commissioning of robust plasma initiation ............................................................................................... 72
2.5.4.2.2 Limiter operation up to 2MA ............................................................................................................................ 73
2.5.4.2.3 Limiter operation up to 3.5 MA ....................................................................................................................... 73
2.5.4.2.4 Divertor operation at 3.5 MA............................................................................................................................ 74
2.5.4.2.5 Time required for obtaining routine divertor operation at 3.5 MA/2.65 T (hydrogen) ......... 74
2.5.4.3 Validation of diagnostic data and demonstration of measurement consistency............................ 75
2.5.4.4 Commission control, interlock and safety systems..................................................................................... 77
2.5.4.4.1 Plasma Control (Conventional control) ....................................................................................................... 77
2.5.4.4.2 Commissioning of the interlock system ....................................................................................................... 79
2.5.4.4.3 Safety System commissioning .......................................................................................................................... 80
2.5.4.4.4 Control commissioning deliverables ............................................................................................................. 80
2.5.4.5 Heating & Current Drive (H&CD) commissioning (ECRH, ICRF) in plasmas up to
7.5 MA/2.65 T ............................................................................................................................................................................... 81
2.5.4.6 Disruption management program in PFPO-1................................................................................................ 84
2.5.4.6.1 Disruption Mitigation System (DMS) commissioning/optimization ............................................... 84
2.5.4.6.1.1 Confirmation of current quench heat load mitigation for 7.5 MA operation ............................ 85
2.5.4.6.1.2 Confirmation of thermal quench heat load mitigation in L-mode ................................................. 85
2.5.4.6.1.3 Confirmation of EM load mitigation scheme for 7.5MA operation ............................................... 86
2.5.4.6.1.4 Confirmation of the injection scheme with maximum margin to RE formation ..................... 86
2.5.4.6.1.5 Confirmation of the margin to RE generation in diverted configuration ................................... 86
2.5.4.6.2 Disruption Load Validation ............................................................................................................................... 87
2.5.4.6.2.1 Confirmation of EM load scaling models.................................................................................................. 87
2.5.4.6.2.2 Confirmation of melt thresholds for thermal loads during the current quench ..................... 87
2.5.4.6.2.3 Confirmation of melt thresholds for the divertor ................................................................................ 87
2.5.4.6.2.4 Confirmation of melt thresholds for the first wall ............................................................................... 87
2.5.4.6.3 Disruption Mitigation System (DMS) trigger generation ..................................................................... 88
2.5.4.6.3.1 Disruption detection ready for high current operation..................................................................... 88
2.5.4.6.3.2 Thermal quench prediction sufficient for L-mode operation ......................................................... 89
2.5.4.6.3.3 Injection sequences ready for high current operation....................................................................... 89
2.5.4.6.3.4 Injection sequences ready for elevated thermal energies ................................................................ 89
2.5.4.7 Development of reliable operation at 7.5 MA/2.65 T ................................................................................ 89
2.5.4.7.1 Establishing 7.5 MA/2.65 T divertor plasmas with heating ................................................................ 89
2.5.4.7.2 Studies of off-axis ECRH at 7.5 MA/2.65 T.................................................................................................. 90
2.5.4.8 Advanced control commissioning in PFPO-1 ................................................................................................ 91
2.5.4.8.1 Plasma kinetic control ......................................................................................................................................... 92
2.5.4.8.2 Magnetohydrodynamics (MHD) and Error Field control ..................................................................... 94
2.5.4.8.3 Supervisory control .............................................................................................................................................. 95
2.5.4.8.4 Advanced control commissioning deliverables for PFPO-1 ................................................................ 96
2.5.4.9 Options for H-mode operation in PFPO-1 ....................................................................................................... 96
2.5.4.9.1 Summary of scenarios ......................................................................................................................................... 96
2.5.4.9.2 Motivation for low field (1.8 T) operation in PFPO-1 ............................................................................ 98
2.5.4.9.3 Details of the experimental plan for H-mode operation in PFPO-1.................................................. 99
2.5.4.9.3.1 First H-mode studies: initial H-mode operation in hydrogen plasma at 5 MA/1.8 T (17days)
99
2.5.4.9.3.2 First ELM control experiments 5 MA/1.8 T (14 days) .................................................................... 100
2.5.4.9.3.3 H-mode in helium at 1.8 T and 5.0 MA (10 days) .............................................................................. 103
2.5.4.9.3.4 Determination of the H-mode threshold in helium plasmas at ~ 2.65 T (3 days)............... 103
2.5.4.9.4 Deliverables for the H-mode operation campaign in PFPO-1 .......................................................... 104
2.5.4.9.5 Assessment of diagnostic capabilities to perform H-mode research during PFPO-1 ............ 106
2.5.4.10 Edge physics and PWI studies in PFPO-1 .................................................................................................. 107

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2.5.4.10.1 Wall Conditioning and cleaning ................................................................................................................. 107
2.5.4.10.1.1 Glow Discharge Cleaning........................................................................................................................... 107
2.5.4.10.1.2 Ion Cyclotron Wall Conditioning/Electron Cyclotron Wall Conditioning............................. 107
2.5.4.10.2 Fuel retention management ........................................................................................................................ 108
2.5.4.10.3 Gas balance in hydrogen ............................................................................................................................... 108
2.5.4.10.3.1 Ammonia formation during nitrogen-seeded discharges ........................................................... 109
2.5.4.10.3.2 Baking studies................................................................................................................................................ 109
2.5.4.10.3.3 Development of a ‘raised strike points’ scenario ............................................................................ 109
2.5.4.10.4 Material migration studies ........................................................................................................................... 110
2.5.4.10.5 Limiter heat load characterization ........................................................................................................... 110
2.5.4.10.6 Heat loads and detachment control during divertor operations ................................................. 113
2.5.5 OPERATIONS PLAN FOR PFPO-2 ................................................................................................................................. 115
2.5.5.1 PFPO-2 Objectives ................................................................................................................................................. 115
2.5.5.2 Plasma restart in PFPO-2 (including influence of Test Blanket Modules) ..................................... 118
2.5.5.2.1 Documenting the effect of the Test Blanket Modules (TBMs) during plasma restart ........... 119
2.5.5.3 Heating & Current Drive (H&CD) commissioning (HNB, DNB, ICRF) .............................................. 119
2.5.5.4 Diagnostics commissioning and validation ................................................................................................. 121
2.5.5.5 Disruption management program in PFPO-2............................................................................................. 122
2.5.5.5.1 Disruption Mitigation System (DMS) commissioning/optimization ............................................ 122
2.5.5.5.1.1 Confirmation of EM load mitigation at 15 MA .................................................................................... 122
2.5.5.5.1.2 Confirmation of thermal quench heat load mitigation in H-mode ............................................. 122
2.5.5.5.2 Disruption Load Validation ............................................................................................................................ 123
2.5.5.5.3 Disruption Mitigation System (DMS) trigger generation .................................................................. 123
2.5.5.5.3.1 Thermal quench prediction ready for FPO .......................................................................................... 123
2.5.5.5.3.2 Injection sequences ready for FPO .......................................................................................................... 123
2.5.5.6 Advanced control commissioning in PFPO-2 ............................................................................................. 124
2.5.5.6.1 Magnetic control ................................................................................................................................................. 124
2.5.5.6.2 Plasma kinetic control ...................................................................................................................................... 124
2.5.5.6.3 Magnetohydrodynamics (MHD) and error field control .................................................................... 126
2.5.5.6.4 Supervisory control ........................................................................................................................................... 127
2.5.5.6.5 Advanced control commissioning deliverables ..................................................................................... 127
2.5.5.7 H-mode Studies in PFPO-2 ................................................................................................................................. 128
2.5.5.7.1 Summary of scenarios ...................................................................................................................................... 128
2.5.5.7.2 Motivation for H-mode scenarios in PFPO-2 .......................................................................................... 131
2.5.5.7.3 Risk to the success of the H-mode studies in PFPO-2 ......................................................................... 132
2.5.5.7.4 Details of experimental plan and deliverables for H-mode scenarios in PFPO-2 .................... 134
2.5.5.7.4.1 H-mode operation to connect with PFPO-1, evaluate/mitigate TBM effects and to extend
H-mode operation at 1.8T (40 days) ................................................................................................................................ 134
2.5.5.7.4.2 Expansion of H-mode operation towards 7.5 MA/2.65 T in hydrogen plasmas (10 days)
136
2.5.5.7.4.3 Expansion of H-mode operation towards 7.5 MA/2.65 T in helium plasmas (15 days) ... 138
2.5.5.7.4.4 Characterization of H-mode operation up to 7.5 MA/2.65 T including ELM control andresolution of scenario integration issues (40 days) .................................................................................................. 139
2.5.5.7.4.5 Demonstration of long-pulse H-mode operation (6 days) ............................................................ 142
2.5.5.7.4.6 Option for low ripple hydrogenic H-modes: H-mode in hydrogen at either 3.0 - 3.3 T and
up to 8.5 or 9.5 MA respectively......................................................................................................................................... 144
2.5.5.7.5 Deliverables for the H-mode scenarios studies in PFPO-2 ............................................................... 145
2.5.5.7.6 Assessment of diagnostic capabilities for performing H-mode research during PFPO-2 .... 146
2.5.5.8 Scenarios for long-pulse operation ................................................................................................................ 146
2.5.5.9 Demonstrate critical system performance (including heat loads) .................................................... 147
2.5.5.9.1 Key elements of critical system performance ........................................................................................ 147
2.5.5.9.2 Critical system performance deliverables in PFPO-2 .......................................................................... 148

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2.5.5.10 Initial studies of current drive efficiency, target q-profile formation and fast particle physics
149
2.5.5.10.1 Assessment of H&CD and diagnostic capabilities for performing current drive studies
during PFPO-2 ........................................................................................................................................................................... 149
2.5.5.10.2 Risk to the initial studies of current drive efficiency, target q-profile formation and fast
particle physics ......................................................................................................................................................................... 150
2.5.5.10.3 Detailed experimental plan for the initial studies of current drive efficiency, target qprofile formation and fast particle physics (16 days) ............................................................................................... 151
2.5.5.10.4 Deliverables for the initial studies of current drive efficiency, target q-profile formationand fast particle physics ........................................................................................................................................................ 153
2.5.5.11 Plasma operation at full technical performance (15 MA/5.3 T) ...................................................... 153
2.5.5.11.1 Requirements .................................................................................................................................................... 154
2.5.5.11.2 Risks to L-mode plasma operation at full technical performance (15 MA/5.3 T) ................ 155
2.5.5.11.3 Detailed experimental plan for the expansion of L-mode plasma operation towards fulltechnical performance (15 MA/5.3 T) (25 days) ........................................................................................................ 155
2.5.5.11.4 Deliverables for the experimental plan for the expansion of L-mode plasma operation
towards full technical performance (15 MA/5.3 T) ................................................................................................... 158
2.5.5.12 Edge physics and PWI studies in PFPO-2 .................................................................................................. 159
2.5.5.12.1 Wall conditioning ............................................................................................................................................. 159
2.5.5.12.2 Fuel retention and material migration ................................................................................................... 159
2.5.5.12.2.1 Deuterium trace experiment ................................................................................................................... 159
2.5.5.12.2.2 Ammonia formation during nitrogen-seeded H-modes ............................................................... 161
2.5.5.12.2.3 Dust generation and characterization ................................................................................................. 162
2.5.5.12.2.4 Post-campaign analysis of Plasma-Facing Components .............................................................. 162
2.5.5.12.3 Heat loads during divertor operations ................................................................................................... 163
2.5.6 SUMMARY OF EXPERIMENTAL ACTIVITIES AND OPERATIONAL TIME IN THE PFPO PHASE .............................. 164
2.6 FUSION POWER OPERATION PHASE (FPO).................................................................................................... 168
2.6.1 OBJECTIVES FOR THE FUSION POWER OPERATION PHASE ..................................................................................... 168
2.6.2 ASSUMPTIONS ................................................................................................................................................................. 172
2.6.3 OPERATION PLAN FOR DEUTERIUM PLASMA EXPERIMENTS ................................................................................... 173
2.6.3.1 Plasma restart in FPO, including H&CD and Diagnostic commissioning ........................................ 173
2.6.3.2 Disruption management program in FPO.................................................................................................... 174
2.6.3.3 Advanced control commissioning in FPO .................................................................................................... 175
2.6.3.3.1 Plasma kinetic control ...................................................................................................................................... 175
2.6.3.3.2 MHD and error field control .......................................................................................................................... 176
2.6.3.3.3 Supervisory control ........................................................................................................................................... 177
2.6.3.3.4 Deliverables for advanced control commissioning in FPO ............................................................... 177
2.6.3.4 Development of deuterium L-mode operation to 15 MA/5.3 T and first assessment of H-mode
access at 2.65 T in D plasmas .............................................................................................................................................. 177
2.6.3.4.1 Summary of scenarios and overall objectives ........................................................................................ 177
2.6.3.4.2 Requirements....................................................................................................................................................... 178
2.6.3.4.3 Risks to the development of deuterium L-mode operation to 15 MA/5.3 T and firstassessment of H-mode access at 2.65 T in D plasmas ............................................................................................... 178
2.6.3.4.4 Detailed experimental plan for the development of deuterium L-mode operation to
15 MA/5.3 T and first assessment of H-mode access at 2.65 T in D plasmas (32 days) ............................. 179
2.6.3.4.4.1 Detailed experimental plan for the development of deuterium L-mode operation to
15 MA/5.3 T (26 days) ........................................................................................................................................................... 179
2.6.3.4.4.2 L-H transition studies in preparation of H-mode operation in D plasmas (6 days) ........... 179
2.6.3.4.5 Diagnostic adequacy for D plasma L-mode scenario studies in FPO ............................................ 180
2.6.3.4.6 Deliverables for the development of deuterium L-mode operation to 15 MA/5.3 T and firstassessment of H-mode access at 2.65 T in D ................................................................................................................. 180
2.6.3.5 Development of deuterium H-mode plasmas towards full performance ....................................... 181

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2.6.3.5.1 Scenarios and overall objectives .................................................................................................................. 181
2.6.3.5.2 Risk to the success of the D H-mode studies in FPO ............................................................................ 182
2.6.3.5.3 Details of experimental plan for D plasma H-mode scenarios in FPO (40 days) ..................... 184
2.6.3.5.4 Diagnostic adequacy for D plasma H-mode scenario studies in FPO............................................ 187
2.6.3.5.5 Deliverables for D plasma H-mode scenario studies in FPO ............................................................ 187
2.6.3.5.6 Progress towards full machine parameters in D H-modes ............................................................... 188
2.6.3.6 Edge physics (including heat loads) and PWI studies in D plasmas ................................................. 189
2.6.3.6.1 Wall conditioning ............................................................................................................................................... 189
2.6.3.6.2 Fuel retention and material migration ...................................................................................................... 190
2.6.3.6.3 Heat loads and detachment control ............................................................................................................ 190
2.6.3.7 Trace Tritium H-mode experiments .............................................................................................................. 192
2.6.3.7.1 Summary of scenarios and objectives........................................................................................................ 192
2.6.3.7.2 Risk to the success of the trace tritium H-mode experiments in FPO .......................................... 192
2.6.3.7.3 Details of experimental plan for the trace tritium H-mode experiments in FPO (12 days) 193
2.6.3.7.4 Diagnostic adequacy for trace tritium H-mode scenario studies in FPO .................................... 194
2.6.3.7.5 Deliverables for the trace tritium H-mode experiments in FPO ..................................................... 194
2.6.3.8 Initial development of hybrid/ advanced scenarios ............................................................................... 194
2.6.3.8.1 Summary of scenarios and overall objectives ........................................................................................ 194
2.6.3.8.2 Risk to the initial development for hybrid/ advanced scenarios ................................................... 195
2.6.3.8.3 Detailed experimental plan for the initial development of hybrid/ advanced scenarios (2040 days) 196
2.6.3.8.4 Deliverables for the initial development of hybrid/ advanced scenarios .................................. 197
2.6.4 OPERATIONS PLAN FOR DEUTERIUM-TRITIUM (DT) PLASMA EXPERIMENTS TOWARDS Q = 10................... 198
2.6.4.1 Plasma scenarios for the experimental program towards Q = 10 operation ................................ 198
2.6.4.2 Optimization of DT plasma scenarios at 7.5 MA/2.65 T (66 days) ................................................... 200
2.6.4.2.1 Experimental plans for the characterization of H-mode access (6 days) ................................... 200
2.6.4.2.2 Experimental plans for the exploration of DT H-mode at 7.5 MA/2.65 T (60 days) .............. 201
2.6.4.2.3 Risks for exploration of DT H-mode at 7.5 MA/2.65 T ....................................................................... 203
2.6.4.2.4 Deliverables from the exploration of DT H-mode at 7.5 MA/2.65 T ............................................. 204
2.6.4.3 Optimization of fusion power in DT plasma scenarios towards 15 MA/5.3 T Q = 10 (~ 50 s)
demonstration (180 days) .................................................................................................................................................... 205
2.6.4.3.1 Risk to the optimization of fusion power in DT plasma scenarios towards 15 MA/5.3 T
Q = 10 (~ 50 s) demonstration ........................................................................................................................................... 211
2.6.4.3.2 Deliverables for the optimization of fusion power in DT plasma scenarios towards
15 MA/5.3 T, Q = 10 (~ 50 s) demonstration ............................................................................................................... 212
2.6.4.4 Extension of the Q = 10 scenario towards long-pulse inductive operation (300 – 500 s) (100days) 2122.6.4.4.1 Risks and deliverable from the experimental phase to extend the Q = 10 scenario towards
long-pulse inductive operation (300 – 500 s) .............................................................................................................. 217
2.6.5 OPERATIONAL PLAN FOR HYBRID/NON-INDUCTIVE DT PLASMA EXPERIMENTS ................................................ 217
2.6.5.1 Summary of scenarios and overall objectives ............................................................................................ 217
2.6.5.2 Risks to operational plan for hybrid/non-inductive DT plasma experiments ............................. 220
2.6.5.3 Experimental plan to develop DT scenarios for long-pulse hybrid and steady-state operation
with high fusion power production .................................................................................................................................. 220
2.6.5.4 Deliverables from the operational plan for hybrid/non-inductive DT plasma experiments . 225
2.6.6 BURNING PLASMA PHYSICS .......................................................................................................................................... 226
2.6.6.1 Opportunities for burning plasma studies in DT plasmas .................................................................... 226
2.6.6.1.1 Fast particle physics and effects of non-axisymmetry ........................................................................ 228
2.6.6.1.2 Self-heating and thermal stability ............................................................................................................... 231
2.6.6.1.3 Macroscopic stability physics and control ............................................................................................... 232
2.6.6.1.4 Multi-scale transport physics ........................................................................................................................ 235
2.6.6.1.5 Physics of the plasma-boundary interface............................................................................................... 235

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2.6.7 SUMMARY OF EXPERIMENTAL ACTIVITIES AND OPERATIONAL TIME IN THE FPO PHASE ................................. 237
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TECHNOLOGY RESEARCH PROGRAM ...................................................................................... 241

3.1 INTRODUCTION.................................................................................................................................................. 241
3.2 TEST BLANKET MODULE (TBM) TESTING PROGRAM IN ITER ................................................................... 242
3.2.1 TEST BLANKET MODULE PROGRAM ........................................................................................................................... 242
3.2.1.1 Background and context for the TBM Program......................................................................................... 242
3.2.1.2 TBM research program accompanying construction .............................................................................. 245
3.2.1.2.1 R&D to be performed on TBMs and associated systems ................................................................... 245
3.2.1.2.2 Effect of TBM ferromagnetic structural material on plasma H-mode performance and fast
ion losses 246
3.2.2 EXPERIMENTAL PROGRAM DURING OPERATION OF THE TBM TESTING PROGRAM IN ITER ........................... 248
3.2.2.1 Overall TBM testing strategy and objectives .............................................................................................. 249
3.2.2.2 Testing objectives for the Electro Magnetic Module (EM-TBM) ........................................................ 251
3.2.2.2.1 Role of the Electro Magnetic Module ......................................................................................................... 251
3.2.2.2.2 General objectives .............................................................................................................................................. 253
3.2.2.2.3 Specific objectives/ measurements for the Lithium-Lead TBM Systems.................................... 253
3.2.2.2.4 Specific objectives/ measurements for the Ceramic-Breeder TBM Systems ............................ 253
3.2.2.2.5 Special requirements for plasma operation scenarios in the PFPO-2 campaign ..................... 254
3.2.2.3 Testing objectives for the Thermal-Neutronic Module (TN-TBM) .................................................... 254
3.2.2.3.1 Role of the Thermal-Neutronic Module .................................................................................................... 254
3.2.2.3.2 General objectives .............................................................................................................................................. 254
3.2.2.3.3 Specific objectives/ measurements for the Lithium-Lead TBM Systems.................................... 255
3.2.2.3.4 Specific objectives/ measurements for the Ceramic-Breeder TBM Systems ............................ 255
3.2.2.3.5 Special requirements for plasma operation scenarios in the FPO-1 campaign........................ 255
3.2.2.4 Testing objectives for the Neutronic-Tritium/Thermo-Mechanic Module (NT/TM-TBM) .... 255
3.2.2.4.1 Role of the Neutronic-Tritium/Thermo-Mechanic Module .............................................................. 255
3.2.2.4.2 General objectives .............................................................................................................................................. 257
3.2.2.4.3 Specific objectives/ measurements for the Lithium-Lead TBM Systems.................................... 257
3.2.2.4.4 Specific objectives/ measurements for the Ceramic-Breeder TBM Systems ............................ 258
3.2.2.4.5 Special requirements for plasma operation scenarios in the FPO-2 campaign........................ 258
3.2.2.5 Testing objectives for the INTegral TBM (INT-TBM).............................................................................. 258
3.2.2.5.1 Role of the INTegral Module .......................................................................................................................... 258
3.2.2.5.2 General objectives .............................................................................................................................................. 259
3.2.2.5.3 Specific objectives/ measurements for the Lithium-Lead TBM Systems.................................... 259
3.2.2.5.4 Specific objectives/ measurements for the Ceramic-Breeder TBM Systems ............................ 259
3.2.2.5.5 Special requirements for plasma operation scenarios in the FPO-3 campaign........................ 259
3.2.2.6 Overall achievements in the TBM Testing Program from PFPO-2 to the FPO-3 campaign –implications for later operations ....................................................................................................................................... 260
3.2.3 SUMMARY AND CONCLUSIONS...................................................................................................................................... 262
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UPGRADE OPTIONS AND REQUIRED R&D ................................................................................ 263

4.1 HEATING AND CURRENT DRIVE SYSTEMS ........................................................................................................ 265
4.1.1 BASELINE H-MODE ACCESS CAPABILITIES ................................................................................................................. 265
4.1.2 STRATEGY TOWARDS H&CD UPGRADES .................................................................................................................... 266
4.1.3 R&D NEEDS TO SUPPORT DECISIONS ON H&CD UPGRADES ................................................................................... 266
4.2 FUELLING AND PUMPING UPGRADES ................................................................................................................ 267
4.3 DIAGNOSTIC UPGRADES BEYOND THE 2016 BASELINE ................................................................................. 269
4.4 UPGRADES OF PLASMA-FACING COMPONENTS AND MATERIALS .................................................................... 271
4.4.1 DIVERTOR ........................................................................................................................................................................ 271
4.4.2 FIRST WALL ..................................................................................................................................................................... 272
4.5 UPGRADES OF DISRUPTION MITIGATION SYSTEMS ......................................................................................... 273

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4.6 UPGRADES TO PLASMA MAGNETIC CONTROL (VERSTICAL STABILIZATION, SHAPE, ERROR FIELDS, RESISTIVEWALL MODES) .............................................................................................................................................................. 274
4.6.1 PLASMA VERTICAL STABILIZATION.............................................................................................................................. 274
4.6.2 CONTROL OF PLASMA CURRENT POSITION AND SHAPE ............................................................................................ 275
4.6.3 CONTROL OF ERROR FIELDS .......................................................................................................................................... 275
4.6.4 CONTROL OF RESISTIVE WALL MODES ...................................................................................................................... 276
4.7 UPGRADES REQUIRED FOR MAINTAINING LONG-PULSE OPERATION ............................................................. 276
4.8 UPGRADES FOR EXPLORATION OF DEMO PLASMA REGIMES AND FOR DEMO TECHNOLOGY DEMONSTRATIONIN ITER ........................................................................................................................................................................ 277
5RESEARCH PROGRAM ACCOMPANYING CONSTRUCTION ................................................ 282

5.1 H-MODE ISSUES ................................................................................................................................................. 282
5.1.1 H-MODE RESEARCH RELATED TO PFPO-1 ................................................................................................................ 283
5.1.2 H-MODE RESEARCH RELATED TO PFPO-2 ................................................................................................................ 283
5.1.3 H-MODE RESEARCH RELATED TO DEUTERIUM AND TRACE T PLASMAS IN FPO.................................................. 284
5.1.4 H-MODE RESEARCH RELATED TO HYBRID AND STEADY-STATE D PLASMAS IN FPO .......................................... 284
5.1.5 H-MODE RESEARCH RELATED TO INDUCTIVE DT PLASMAS IN FPO...................................................................... 284
5.1.6 H-MODE RESEARCH RELATED TO HYBRID AND STEADY-STATE DT PLASMAS IN FPO ....................................... 285
5.2 PLASMA-WALL INTERACTION ISSUES ............................................................................................................... 285
5.3 MHD STABILITY ISSUES .................................................................................................................................... 288
5.3.1 DISRUPTION CHARACTERIZATION, PREDICTION AND MITIGATION ........................................................................ 288
5.3.2 SAWTOOTH CONTROL .................................................................................................................................................... 290
5.3.3 ELM CONTROL ................................................................................................................................................................ 291
5.3.4 NEOCLASSICAL TEARING MODE CONTROL .................................................................................................................. 292
5.3.5 ERROR FIELD CONTROL ................................................................................................................................................. 293
5.3.6 RESISTIVE WALL MODE CONTROL .............................................................................................................................. 294
5.4 SCENARIO DEVELOPMENT ISSUES ..................................................................................................................... 294
5.4.1 BREAKDOWN AND BURN-THROUGH ............................................................................................................................ 295
5.4.2 CURRENT RAMP-UP ........................................................................................................................................................ 296
5.4.3 CURRENT FLAT-TOP ....................................................................................................................................................... 296
5.4.4 PLASMA TERMINATION .................................................................................................................................................. 297
5.4.5 IMPACT OF SCENARIO DEVELOPMENT ISSUES ON OPERATION ................................................................................ 298
REFERENCES ............................................................................................................................................ 300
APPENDIX A: USE OF DEUTERIUM SEEDING TO CHARACTERIZE FUEL RETENTION
DURING PFPO-2 PHASE .......................................................................................................................... 312
APPENDIX B: L-H THRESHOLD AND TOROIDAL FIELD RIPPLE EFFECTS ON H-MODES 317
APPENDIX C: HEAT LOAD MANAGEMENT ..................................................................................... 326
APPENDIX D: REFERENCE PULSE ...................................................................................................... 333
APPENDIX E: NON-INDUCTIVE SCENARIOS WITH/WITHOUT LHCD ..................................... 334
APPENDIX F: PHYSICS ANALYSIS AND RISKS OF 5 MA/1.8 T SCENARIOS ............................ 343
APPENDIX G: HEATING AND CURRENT DRIVE STAGING.......................................................... 356
APPENDIX H: DIAGNOSTIC STAGING ............................................................................................... 379
APPENDIX I: TRITIUM AVAILABILITY ............................................................................................. 401
APPENDIX J: RESEARCH PLAN RISK REGISTER ........................................................................... 403

7

ITR-18-003

Acronyms
ACCC
AE

Active Compensation and Correction Coils
Alfvén Eigenmodes
Tokamak operating between 1991 and 2016 at the Massachusetts
Alcator C-Mod
Institute of Technology
ASDEX-Upgrade Tokamak developed at the Max-Planck-Institute for Plasma Physics
BES
Beam Emission Spectroscopy
BLV
Beam Line Vessel
BSV
Beam Source Vessel
CC
Correction Coils
CCB
Change Control Board
CCWS
Component Cooling Water System
CEA
Commissariat à l’énergie atomiqueCI
Coherence Imaging
CIS
Central Interlock System
CODAC
Control, Data Access and Communication
CN-DA
China Domestic Agency
CQ
Current Quench
CS
Central Solenoid
CSs
Cooling Systems
CSS
Central Safety System
CTS
Collective Thomson Scattering
CX
Charge-Exchange
CXRS
Charge Exchange Recombination Spectroscopy
DA
Domestic Agency
DBC
Disruption Budget Consumption
DCLL
Dual-Coolant Lithium-Lead
DEFC
Dynamic Error Field Correction
DEMO
Demonstration Fusion Reactor (next step after ITER)
DIII-D
Tokamak developed in the 1980s by General Atomics
DIM
Divertor Impurity Monitor
DIP
Density Interferometer Polarimeter
DMS
Disruption Mitigation System
DNB
Diagnostic Neutral BeamDT
Deuterium-Tritium
EAST
Tokamak developed in Hefei, China
EC
European Commission
EC
Electron Cyclotron
ECCD
Electron Cyclotron Current Drive
ECE
Electron Cyclotron Emission (Diagnostic)
ECH&CDElectron Cyclotron for Heating & Current Drive
ECR
Electron Cyclotron Resonance
ECRH
Electron Cyclotron Resonance Heating
8

ITR-18-003

ECWC
EDA
EHF
EL
ELM
EM
EM-TBM
EPED
EPM
ERM/KMS
EU-DA
ETS
F4E
FC
FDR
FI
FICX
FILD
FIR
FOCS
FP
FPO
FPPC
FS
FW
FWP
GDC
GIM
GIS
H&CD
HCCB
HCCR
HCLL
HCPB
HFS
HNB
HQ
HRNS
HUST
HV
HVPS
I&C
IC
ICH&CD
ICRF

Electron Cyclotron Wall Conditioning
Engineering Design Activities
Enhanced Heat Flux
Equatorial Launcher
Edge Localized Mode
Electro-Magnetic
Electro Magnetic module
Name of model describing Edge Pedestal
Fast Particle Modes
Ecole Royale Militaire / Koninklijke Militaire School
European Union Domestic Agency
Edge Thomson Scattering
Fusion fur Energy
Fission Chambers
Final Design Review
Ferritic Inserts
Fast Ion Charge Exchange
Fast Ion Loss Detector
Fast Infrared
Fibre Optic Current Sensor
First Plasma
Fusion Power Operation Phase
First Plasma Protection Component
Faraday ScreenFirst WallFirst Wall Panels
Glow discharge cleaning
Gas Injection Module
Gas Injection System
Heating & Current Drive
Helium-Cooled Ceramic Breeder
Helium-Cooled Ceramic Breeder Graphite Reflector
Helium-Cooled Lithium-Lead
Helium-Cooled Pebble Beds
High Field Side
Heating Neutral Beam
ITER Headquarters
High Resolution Neutron Spectrometer
Huazong University of Science and Technology
High Voltage
High Voltage Power Supply
Instrumentation and Control
Ion Cyclotron
Ion Cyclotron for Heating & Current Drive
Ion Cyclotron Range of Frequencies
9

ITR-18-003

ICRH
ICWC
IMAS
INT-TBM
IO
IO-CT
IO-DA
IPP
IPR
IR
IRFM
IRP
ISS
ISS
ITB
ITPA
IVT
JA-DA
JET
JET-ILW
JT-60
KSTAR
LB
LCFS
LFS
LH
LHCD
LHH&CD
LIF
LLCB
LLNL
LM
LM
LOS
LSM
MAST
MD
MGI
MHD
MIMO
MOU
MPD
MSE
NB

Ion Cyclotron Resonance Heating
Ion Cyclotron Wall Conditioning
Integrated Modelling and Analysis Suite
INTegral TBM
ITER Organization
ITER Organization Central TeamITER Organisation Domestic Agency
Max Planck Institute of Plasma Physics
Institute for Plasma Research
InfraredInstitut de la Recherche sur la Fusion par confinement Magnétique
ITER Research Plan
Isotope Separation System
Interspace Support StructureInternal Transporter Barrier
International Tokamak Physics Activity
In-Vessel Transporter
Japan Domestic Agency
Joint European Torus
JET ITER-Like Wall (material mix)
Tokamak developed in 1990s by the Japan Atomic Energy ResearchInstituteTokamak developed in Korea by the National Fusion Research InstituteLiquid Breeders
Last Closed Flux Surface
Low Field Side
Lower Hybrid
Lower Hybrid Current Drive
Lower Hybrid Heating and Current Drive
Laser Induced Fluorescence
Lithium-Lead Ceramic Breeder
Lawrence Livermore National Laboratory
Liquid Metal
Locked Mode
Line-Of-Sight
Lower Steering Mirror
Tokamak developed in the UK by the Culham Centre for Fusion Energy
Maintenance Detritiation
Massive gas injection
Magnetohydrodynamics
Multiple-Input Multiple-Output
Matching Optics Unit
Multi-Purpose Deployer
Motional Stark Effect
Neutral Beam
10

ITR-18-003

NBCD
NBI
NBTF
NFRI
NN
NRCKI
NSTX
NTM
NT/TM-TBM
NTV
NWL
PA
PAM
PCR
PCS
PCSS
PCSSP
PF
PFC
PFPO-1PFPO-2
PMS
PMT
PID
PIS
PP
PPEN
PPPL
PPTF
PS
PVTc
PWI
QST
R&D
RAFM
RAMI
RE
RES
RF
RF-DA
RGA
RGRS
RH

Neutral Beam Current Drive
Neutral Beam Injection
Neutral Beam Test Facility
National Fusion Research Institute (KR)Neural Network
National Research Center “Kurchatov Institute”
National Spherical Torus Experiment magnetic fusion device developedby the Princeton Plasma Physics Laboratory
Neoclassical Tearing Modes
Thermo-Mechanic module
Neoclassical Toroidal Viscosity
Neutron Wall Load
Procurement Arrangement
Passive-Active Multi-Junction
Project Change Requests
Plasma Control System
Port Cell Support Structure
Plasma Control System Simulation Platform
Poloidal Field
Plasma Facing Component
Pre-Fusion Power Operation 1
Pre-Fusion Power Operation 2
Passive Magnetic Shield
Photomultiplier Tube
Proportional-Integral-Derivative
Plant Interlock Systems
Port Plug
Pulsed Power Electrical Networks
Princeton Plasma Physics Lab
Port Plug Test Facility
Power Supply
Pressure-Volume-Temperature-composition
Plasma Wall Interaction
National Institutes for Quantum and Radiological Science andTechnology
Research and Development
Reduced-Activation Ferritic/Martensitic
Reliability, Availability, Maintainability & Inspectability
Runaway Electrons
RE Suppression
Radio-Frequency
Russia Domestic Agency
Residual Gas Analyzers
Radial Gamma-Ray Spectrometer
Remote Handling
11

ITR-18-003

RMP
RMS
RNC
RSAE
RVTL
RWM
SB
SCM
SOL
SPA
SPI
SSEN
STAC
SWIP
TAE
TBM
TBMA
TMB-PC
TBS
TCV
TCWS
TF
TIP
TL
TL
TLM
TNS
TQ
UKAEA
UL
USM
VDE
VGRS
VNC
VS
VSWR
VUV
VV
WCCB
XRCS

Resonant Magnetic Perturbations
Root Mean Square
Radial Neutron Camera
Reversed Shear Alfvén Eigenmodes
Removable Vacuum Transmission Line
Resistive Wall Mode
Solid Breeders
Superconducting Magnet
Scrape-Off Layer
Single-Pass Absorption
Shattered Pellet Injection
Steady State Electrical Networks
Science and Technology Advisory Committee
Southwestern Institute of Physics
Toroidal Alfvén Eigenmodes
Test Blanket Module
TBM Arrangement
TBM Program Committee
Test Blanket Systems
Tokamak à Configuration Variable (Swiss Tokamak)
Torus Cooling Water System
Toroidal Field
Toroidal Interferometer Polarimeter
Member Leader
Transmission Line
Thermal Load Mitigation
Tangential Neutron Spectrometer
Thermal Quench
UK Atomic Energy Authority
Upper Launcher
Upper Steering Mirror
Vertical Displacement Events
Vertical Gamma Ray Spectrometer
Vertical Neutron Camera
Vertical Stabilisation
Voltage Standing Wave Ratio
Vacuum Ultra Violet
Vacuum Vessel
Water-Cooled Ceramic Breeder
X-ray Crystal Spectroscopy

12

ITR-18-003

Acknowledgements
The ITER Research Plan has been developed over a period of more than 10 years through an
extensive collaboration between staff of the ITER Organization and experts from the Members’
Domestic Agencies and fusion communities. Contributions from the following individuals to the
preparation of the present version of the Plan are gratefully acknowledged:

P Andrew, IO
I Bandyopadhayay, IN DA (IN)R Barnsley, IO
B Beaumont, IO
L Bertalot, IO
D Boilson, IO
X Bonnin, IO
C Bourdelle, CEA/IRFM (EU)D Campbell, IO
A Chakraborty, IPR (IN)D Chandra, IPR (IN)M Clough, IO
C Darbos, IO
M De Bock, IO
G De Temmerman, IO
P de Vries, IO
D Douai, CEA/IRFM (EU)X Duan, SWIP (CN)J Elbez-Uzan, IO
M Fenstermacher, LLNL (US)A Fossen, IO
L Giancarli, IO
J Graceffa, IO
M Graham, IO
Y Gribov, IO
J Gunn, CEA/IRFM (EU)M Henderson, IO
S-H Hong, NFRI (KO)G Huijsmans, EC (EU)Y-K In, NFRI (KO)A Isayama, QST (JA)S Kaye, PPPL (US)Y Kazakov, ERM/KMS (EU)F Kazarian, IO
A Khan, IO
S-H Kim, IO
N Kirneva, NRCKI (RF)M Kocan, IO
S Konovalov, NRCKI (RF)Ph Lamalle, IO
M Lehnen, IO
V Leonov, NRCKI (RF)E Lerche, ERM/KMS (EU)S Lisgo, IO
X Litaudon, EUROfusion (EU)A Loarte, IO
T Luce, IO&GA (US)R Maingi, PPPL (US)S Maruyama, IO
Y-S Na, Seoul Nat Uni (KO)R Neu, IPP (EU)I Nunes, IST (EU)Y-K Oh, NFRI (KO)S Pinches, IO
R Pitts, IO
A Polevoi, IO
R Reichle, IO
F Rimini, UKAEA (EU)A Romannikov, NRCKI (RF)R Rotella, IO

13

S Sabbagh, Columbia Uni (US)V Safronov, RF DA (RF)G Saibene, F4E (EU)R Sartori, F4E (EU)M Schneider, IO
K Schmid, IPP (EU)B Schunke, IO
M Singh, IPR (IN)G Sips, EC (EU)A Sirinelli, IO
J Snipes, IO
J Stober, IPP (EU)T Suzuki, QST (JA)L Svensson, IO
Y Takase, Uni Tokyo (JA)H Takenaga, QST (JA)E Tsitrone, CEA/IRFM (EU)V Udintsev, IO
H Urano, QST (JA)J van der Laan, IO
D Van Eester, ERM/KMS (EU)G Vayakis, IO
E Veshchev, IO
M Walsh, IO
C Watts, IO
S Willms, IO
S-W Yoon, NFRI (KO)
L Zabeo, IO
G Zhuang, HUST EDU (CN)ITR-18-003

Executive Summary
The ITER Research Plan (IRP) has been developed to provide a guide to the overall research
activities which should be undertaken within the framework of the ITER Project. When complete,
the IRP should encompass both physics and technology research during ITER construction andoperation (though technology R&D associated with the procurement packages formally falls within
the Project Plan and Resource Estimate, rather than the Research Plan). The revised version
presented here focusses on physics R&D and on the R&D and testing program of the (tritium
breeding) Test Blanket Modules (TBM) adapted to the revised ITER project schedule developed
during 2015 - 2016 within the Staged Approach. Ultimately, this document should aim toencompass activities such as operational R&D on heating and current drive and diagnostic systems,
and operational R&D on the broader aspects of fusion technology associated with major ITER
subsystems, such as superconducting magnets, remote handling, tritium handling technology etc. inorder to provide a complete overview of the research scope and plans of the ITER project.
This version of the Research Plan is a provisional version of the baseline documentation that has
been developed in response to the complete revision of the ITER project schedule within theframework of the Staged Approach, undertaken in 2015 – 2016 and reflects the ITER project
strategy as of December 2017. The Staged Approach foresees First Plasma in December 2025,
which is succeeded by a progressive upgrade of the capabilities of the ITER tokamak and facility
interleaved with two periods of system ‘commissioning with plasma’ and experimental plasmastudies in H and He plasmas. Completion of construction activities is scheduled for early 2035 andthe transition to experiments in D/DT plasmas is planned for December 2035, with trace tritium
experiments likely in early 2036 and a gradual transition to fusion power production over the next
12 – 15 months of experimental studies, leading to an initial demonstration of several hundred
megawatts of fusion power production for several tens of seconds. In subsequent experimental
campaigns in Deuterium Tritium (DT) plasmas, planned on a two-yearly cycle, the experimental
basis for achieving the principal scientific mission goals of the ITER project are developed: ademonstration of Q ≥ 10 for burn durations of 300 – 500 s and the development of long-pulse, noninductive scenarios aiming at maintaining Q ~ 5 for periods of up to 3000 s. The plan incorporates
elements detailed below to adapt the baseline version of the Research Plan (v2.2, November 2009)
to the Staged Approach:(i) Incorporation of a first divertor with all-tungsten plasma-facing components, installed
during Assembly Phase II, following First Plasma;(ii) Development of the Research Plan around a progressive increase in the installed Heating &Current Drive (H&CD) power towards the baseline capability of 73 MW and a staged
installation of Diagnostics matching the measurement requirements of the Research Plan;(iii) Exploration of options for studies of H-modes at lower toroidal magnetic field (1.8 T) and in
H (as well as He), which was not previously considered feasible;(iv) Strengthening of the development of a disruption management program to characterizedisruption loads, and to implement effective disruption prediction, avoidance and mitigation;(v) A significantly more extensive program of studies on plasma-wall interactions tocharacterize issues such as erosion, transport and redeposition of plasma-facing materials,
fuel retention and removal, and degradation of plasma-facing materials under plasma
exposure;

14

ITR-18-003(vi) Start of commissioning of the Tritium Plant with tritium in 2033, during the second
experimental campaign in H/He plasmas;(vii) Transition to experiments in D/DT plasmas in December 2035, with the start of trace
tritium experiments in early 2036, as a result of the accelerated commissioning of the
Tritium Plant;(viii) Progressive transition to full DT operation during 2036, with an expanding program of DT
operation which aims to demonstrate fusion power production at a level of several hundred
megawatts for several tens of seconds by the end of the first experimental campaign in
D/DT plasmas in early 2037;(ix) Development of the experimental program during the second and third DT campaigns
towards the achievement of ITER’s fusion power, fusion gain and long burn duration
mission goals;(x)Adaptation of the evolution of the Test Blanket Module testing program to the revised
ITER schedule and experimental plan.

The structure of the Research Program has been analyzed within 4 broad areas: the experimental
program during operations, the TBM development and testing program (as the first element of theTechnology research program, still to be developed in detail), the possible upgrade program for thedevice’s capabilities and the research program during construction. The issues which will be
addressed within these areas can be summarized as follows:
Experimental Program during Operations
Within the framework of the Staged Approach, the ITER Research Plan has analyzed the physics
research and plasma operation requirements from First Plasma, in December 2025, to the end of the
3rd experimental campaign in DT plasmas, expected to close (according to current assumptions
about the cycle of ITER operations during the fusion power operation phase (FPO)) in spring 2041,
at which time the long-pulse (300 – 500 s) Q ≥ 10 goal of the ITER mission should have beenachieved at power levels of ~500 MW, and significant progress made towards the achievement of
Q ~ 5 for burn durations of ~3000 s in non-inductive plasmas. In this version, the IRP has
developed an experimental program to fully commission plasma operation of the device, auxiliary
and plant systems, and to achieve the level of DT plasma performance required satisfy the fusion
performance goals of the project.
The principal aims and activities encompassed by the several operational phases spanning thisperiod of ITER operation are summarized below.
Pre-Fusion Power Operation (PFPO)
 This operational phase will consist of two periods of plasma operation, a total of 37 monthsof experiments in H and He plasmas, interleaved with Assembly and Integrated
Commissioning activities, to bring the tokamak, its auxiliary systems and facility plantsystems up to their full performance level and capabilities; During this phase H and He plasma scenarios will be developed to allow the fullcommissioning of all tokamak sub-systems (except those involving the use of D or T) withplasma; It is planned that all baseline H&CD systems will be tested with plasma operation andbrought up to the specified level of input power; options for dual-frequency operation of the15

ITR-18-003

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














Electron Cyclotron Resonance Heating (ECRH) system and early operation of one Ion
Cyclotron Rage of Frequencies (ICRF) antenna are being explored; Diagnostics, fuelling,
control, protection and safety-related systems will also be commissioned;
The Vertical Stability coils, Error Field Correction Coils and Edge Localized Mode (ELM)
control coils will be commissioned and exploited routinely;
The level of electromagnetic and thermal loads associated with disruptions and Vertical
Displacement Event (VDEs) will be determined; disruption avoidance and mitigation
techniques will be implemented and applied routinely;
In-vessel components will be tested to the limits of the available thermal loads, andtechniques for mitigation of heat loads will be developed;
Methodologies for the characterization of fuel retention and dust production will be
implemented and techniques for hydrogen and dust removal will be demonstrated; analysis
indicates that a short campaign with low-level deuterium seeding of hydrogen plasmas could
provide accurate measurements of fuel retention if any regulatory issues can be resolved;
Initial H-mode operation will be established, H-mode behaviour characterized and ELM
control techniques established; H-mode operation in He appears more accessible, but
options for studying H-modes in H are being explored;
A key milestone for the Pre-Fusion Power Operation phase (PFPO) will be the demonstrationof L-mode plasmas at the full technical capability of the device (15 MA/5.3 T);
Plasma pulse length, including that in plasmas at reduced parameters, will likely be limitedby the operational time available to extend the pulse length for H&CD systems (andpossibly also other auxiliary systems) to long-pulse operation; a minimal target of ~50 s has
been established, but scenarios are available for plasma pulse lengths of 100 – 1000 s if
sufficient experimental time is available;
Initial studies of current drive efficiency and current profile formation will be performed;
Diagnostic data will be validated and the consistency of the various measurements will bedemonstrated;
The interpretive software chain for the reconstruction of the plasma evolution will be refinedand validated;
During the PFPO-2 phase, the Test Blanket Module (TBM) program will test Electro
Magnetic modules.

Deuterium Operation (Fusion Power Operation, Campaign 1)
 Certain ‘commissioning with plasma’ activities will be undertaken during initial deuterium
operation, involving recommissioning of H&CD systems in deuterium (e.g. Neutral Beam
(NB) injection with 1 MeV D0), commissioning of any nuclear measurement systems (e.g.
neutron detectors) and newly installed Diagnostics, commissioning of fuelling systems with
deuterium (and eventually tritium); commissioning of additional control and interlock
systems in preparation for nuclear operation;
 Disruption detection, avoidance and mitigation will be adapted to the highest parameters
achievable in D plasmas;
 Plasma scenarios will be redeveloped up to the full technical parameters (15 MA/5.3 T) inD;

16ITR-18-003











Once commissioning of such systems is complete, a program of H-mode studies, including
ELM control and divertor power load control, will be implemented, progressing to thehighest parameters where robust H-mode operation can be sustained (possibly
7.5 MA/2.65 T);
Development of reliable Neo-classical Tearing Mode (NTM) control in high confinement D
H-mode plasmas, if not previously implemented in the PFPO program;
The research program will then focus on characterizing the physics of D plasmas at the
ITER scale in ohmic, L- and H-mode plasmas;
Low concentration (trace) T experiments in D L- and H-mode plasmas will assess Ttransport and T fuelling requirements, and initiate the transition to DT experiments;
Characterization of key aspects of plasma-wall interactions in the ITER environment in D
plasmas will be performed;
Preparatory studies of hybrid/non-inductive scenarios in D plasmas will be undertaken;
Preparatory studies of ‘burning plasma’ issues (e.g. helium exhaust, fuel mixture control in
trace T, etc.) will also be carried out to prepare control techniques for DT plasmas;
In this operational phase (extending into the initial period of DT operation), the TBM
program will test Thermal Neutronic modules.

Deuterium-Tritium Operation (Fusion Power Operation, Campaign 1)
 The initial DT experimental program will focus on developing the inductive ELMy
H-mode scenario which provides the basis for achieving ITER’s principal fusion power(500 MW) and fusion gain (Q ≥ 10) mission;
 Additional experimental ‘commissioning’ will be required in order to develop all of thenecessary measurement systems (e.g. fusion product measurements) and control techniques
to support integrated scenarios at high fusion power; in addition, ICRF heating scenarios for
DT plasmas will be commissioned; As the program progresses and the level of fusion power rises, disruption detection,
avoidance and mitigation techniques will have to be adapted;
 DT H-modes are likely to be developed around 7.5 MA/2.65 T with durations of several tensof seconds (~50 s), incorporating reliable ELM control, NTM control and radiative divertor
operation; The program will thereafter emphasize the rapid development of inductive operation,
culminating in 12 - 15 MA/5.3 T DT H-mode operation with full feedback control at fusion
powers of 200-300 MW for at least 50 s;
 Experiments would explore optimization of fuel mixture control, burn control, helium
exhaust, plasma-facing component protection in DT plasmas;
 These plasmas would allow characterization of physics processes in DT ohmic and L-mode
scenarios and initiate studies of burning plasma physics at Q ≥ 5 in H-mode; If time permits, the development would continue towards optimization of plasma parameters
to maximize Q, with the goal of reaching Q = 10 for ~50 s as rapidly as possible.

17

ITR-18-003

Deuterium-Tritium Operation (Fusion Power Operation, Campaigns 2 and 3)
 Inductive scenario development will focus on achieving ITER’s primary fusion performancegoal of Q ≥ 10 at fusion power levels of 400 – 500 MW and at plasma burn durations of
300 - 500 s;
 Studies to characterize key aspects of PWI in the ITER environment in high fusion power/
high gain plasmas will be undertaken; in particular, demonstration of satisfactory tritium
accounting, retention measurement and removal techniques will be an essential aspect of the
program; Optimization of fusion power with Q > 10 during stationary or transient inductive conditions
will be explored and plasma behaviour characterized in these conditions; Development of non-inductive current drive and current profile control in DT plasmas will
support the exploration of both hybrid scenarios and fully non-inductive plasma regimes; It will be important to identify needs for upgrades to improve fusion performance or to
improve the potential for non-inductive operation (e.g. ELM control coil power supplies for
Resistive Wall Mode (RWM) control, if it has not been possible to do so at an earlier stageof the program); The program developed towards very long pulse, non-inductive operation foresees theestablishment of DT plasma scenarios with current profile control and improved H-mode
confinement having a burn duration of ~1000 s and several hundred megawatts of fusion
power, based on the hybrid /improved H-mode scenario at q95 ~ 4;
 Non-inductive plasmas with the longest sustainable pulse duration (> 1000 s) and moderate
fusion performance (Q ~ 2) would then be developed as a basis for meeting the noninductive mission goal; In parallel, the program would aim to optimize fusion gain and pulse duration in the mostpromising long-pulse scenario(s) to establish plasmas with duration beyond 1000 s andQ ≥ 5;
 This development would provide a plasma scenario allowing characterization of key aspectsof Plasma-Wall-Interaction (PWI) in the ITER environment over very long durations (e.g.
impact on fuelling, erosion/redeposition, fuel retention, etc.);
 The mission goal of demonstrating Q ~ 5 in fully non-inductive DT plasmas and extending
this to the longest pulses which the ITER facility can support (nominally ~3000 s) would
then be a major focus of a program which might require further experimental campaigns torealize fully; The establishment of plasma regimes in which -particle heating is dominant, the key novel
feature of ITER plasmas, will provide access to a wide range of burning plasma physics
studies which will be integrated into the programs aimed at achieving the fusion
performance milestones in both inductive and non-inductive plasmas;
 During the 2nd experimental campaign in DT, the TBM program will test a NeutronicTritium/Thermo-Mechanic module, while in the 3rd experimental period an INTegral
module will be tested; The DT research program pursued during these initial FPO campaigns would also provide
an opportunity to identify options for further upgrades to prepare longer-term Demotargeted studies.

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TBM Development and Testing Program
The TBM development program involves the design and construction of up to 6 TBM systems(TBMs plus all ancillary systems required for the operation and testing of the TBMs) by Members’
teams, together with the design and construction, by the IO, of the required infrastructure to host the
TBM systems within the ITER facility and to ensure all interfaces to ITER plant systems. The
major elements of R&D required to support the development of the TBM systems have been
identified. A testing program for the TBMs, to be implemented during operations, has also beendeveloped and integrated into the experimental program. It is expected that the TBM testing
program will be undertaken without interference to the physics research program during the majorpart of this period. As high fusion power, very long burn duration plasmas are developed in the 2nd
and 3rd DT campaigns, it is likely that continuous, repetitive operation of ITER will be required forshort periods (~6 days) to enable specific performance tests of TBMs and their associated systems.
The structure of the TBM testing program foreseen at present is based on the successive installationand testing of 4 variants of each of the up to 6 TBM concepts being developed by the Members:
 During the PFPO-2 phase, the TBM program will test Electro Magnetic modules, which canallow the assessment of the overall functionality of the TBM systems, including theiroperation and control, and provide confirmation of the thermal behaviour of the TBM firstwalls in the tokamak environment; within the scientific program, studies of the impact oferror fields and toroidal field ripple due to the TBM ferromagnetic materials will be
undertaken and mitigation measures implemented where possible;
 During the FPO-1 campaign, encompassing both the D phase and the initial period of DT
operation, the TBM program will test Thermal Neutronic modules, which are used toinvestigate neutronic response and to confirm the thermal behaviour of the TBMs;
 Neutronic-Tritium/Thermo-Mechanic modules will be tested during the 2nd experimental
campaign in DT; this allows investigation of the thermo-mechanical behaviour at relevanttemperatures and also permits the nuclear response to a Demo-relevant 14.1 MeV neutron
flux to be determined, focussing on the actual tritium breeding and extraction functions; In the 3rd DT experimental period, an INTegral module will be tested in order to obtain
information about tritium management and heat extraction in fully representative Demobreeding blanket conditions, including reproducible plasma performance; the experimental
program will provide support to the testing of the INTegral module by preparing appropriate
plasma pulses and pulse sequences.
The general objectives of the testing programs relating to each of the 4 stages outlined above andspecific objectives of the R&D on different TBM concepts (liquid breeder/ ceramic breeder) havebeen identified. Guidelines for desirable modes of plasma operation to facilitate the TBM testing
program have also been developed.
Upgrade Program
During the period of ITER construction and of initial plasma operation, it is anticipated that
proposals for upgrades to the H&CD, diagnostic, fuelling and control systems will be developed
and more concrete plans for changes to the ITER plasma-facing components will be put in place. As
funding becomes available for the implementation of the highest priority upgrades, the operational
program will need to be revised to accommodate the additional flexibility and capabilities whichsuch upgrades will bring. This is likely to be even more relevant for the planning for the longerterm operation, focussing on optimization of fusion performance, extended pulse length, developing
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a detailed understanding of burning plasma physics, and the investigation of Demo-relevant scienceand technology issues, which can be expected to benefit most from the upgrade program.
Research Program during Construction
The major focus of the physics research program accompanying construction will be to enhance thephysics and technology basis for operation. The key activities which will need to be addressed inorder to prepare for efficient experimental exploitation of the ITER facility have been identified as: H-mode access and characteristics, including core and pedestal transport in transient andstationary conditions, ELM mitigation, influence of 3-D fields;
 Plasma-wall interaction issues, including tritium retention and removal, dust production andremoval, heat loads and power handling techniques, and the implications of these
phenomena for Plasma Facing Component (PFC) lifetime; Magnetohydrodynamics(MHD)
stabilityandcontrol,
includingeffective
disruption/runaway electron mitigation measures, and control of sawteeth, neoclassicaltearing modes, resistive wall modes and error field induced modes;
 Development of a comprehensive integrated modelling capability to support plasma
operations and research; Scenario development, encompassing the development of all plasma scenario phases (rampup, entry and exit from burn and ramp-down) and the necessary control techniques for therange of reference scenarios foreseen.
Appendices
The ITER Research Plan also contains several Appendices which deal with aspects of physics,
phasing of auxiliary systems within the Staged Approach, ‘background’ issues such as Tritium
Availability and risks to the successful implementation of the research program: Appendix A: analysis of the case for deuterium seeding of hydrogen plasmas for improved
characterization of fuel retention during the PFPO phase.
 Appendix B: discussion of H-mode power threshold and the influence of Toroidal Field(TF) ripple on H-mode performance as they impact the Research Plan.
 Appendix C: discussion of heat load management and the evolution of the expected heatloads to PFCs during the course of the Research Plan.
 Appendix D: definition of the requirements of a reference pulse which could be run
regularly within the experimental program to characterize gradual changes in the plasma
operating environment – of particular relevance to PWI studies.
 Appendix E: a review of modelling studies of the options for very long-pulse inductive andnon-inductive scenarios in ITER and, in particular, the importance of Lower Hybrid Current
Drive (LHCD) in such scenarios.
 Appendix F: discussion of the results of physics analysis undertaken in relation to the option
for operation at 5 MA/1.8 T and of the physics risks associated with these plasma scenarios.
 Appendix G: discussion of the phasing of the H&CD systems during the Staged Approachand the current preparations for possible H&CD upgrades.
 Appendix H: discussion of the phasing of the plasma diagnostic systems during the Staged
Approach and its impact on measurement requirements.

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

Appendix I: brief analysis of expected availability of tritium from ‘commercial’ sources for
the ITER DT program.
Appendix J: Risk Register for the most significant physics risks to the successful
implementation of the Research Plan.

Future Development of the Research Plan
It is intended that the ITER Research Plan should be subject to continuous development during the
lifetime of the Project and, in time, should provide an instrument for managing and guiding the
Project’s scientific and technological research activities. Key analysis which should be pursued in
the medium-term to expand the present document and integrate the scientific research program
more completely with the overall Project development includes:
 Further detailing of the research program during construction with clear links to the ongoing
fusion R&D programs in the Members’ fusion communities; Detailing of elements of the Integrated Commissioning program to develop a more complete
plan for the transition to plasma operation;
 Assimilation of more detailed analysis of plasma commissioning requirements for H&CD
and diagnostic systems, including off-line commissioning requirements in preparation forcommissioning with plasma; Elaboration of the program of burning plasma physics research, with a more detailed
development of the major elements of this program; Development of a technology research program for the operations phase encompassing
R&D activities on tokamak, auxiliary and plant systems (in addition to the TBM testing
program discussed here) to provide an extensive technology basis for the design of future
fusion power plants;
 Deeper analysis of the potential impact of possible upgrade paths on the first 10 years of theexperimental program.

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1 Introduction
1.1 Objectives
The Project Specification [ITER_D_2DY7NG, 2010] defines the overall programmatic objective of
the ITER project as, ‘to demonstrate the scientific and technological feasibility of fusion energy forpeaceful purposes’. It also sets out a series of scientific and technical goals which should besatisfied to achieve this objective. The scientific goals, defined in terms of plasma fusion
performance, are:
 The device should achieve extended burn in inductively driven plasmas with the ratio of
fusion power to auxiliary heating power, Q, of at least 10 (Q ≥ 10) for a range of operating
scenarios and with a duration sufficient to achieve stationary conditions on the timescales
characteristic of plasma processes.
 The device should aim at demonstrating steady-state operation using non-inductive current
drive with the ratio of fusion power to input power for current drive of at least 5.
 In addition, the possibility of controlled ignition should not be precluded.
The technical goals, aiming to provide much of the technological basis for the design of future
fusion power plants capable of generating electricity, are:
 The device should demonstrate the availability and integration of technologies essential for afusion reactor (such as superconducting magnets and remote maintenance).
 The device should test components for a future reactor (such as systems to exhaust power
and particles from the plasma).
 The device should test tritium breeding module concepts that would lead in a future reactor
to tritium self-sufficiency, the extraction of high grade heat and electricity production.
The primary aim of the ITER Research Plan (IRP), therefore, is to define the plan of research anddevelopment and of facility exploitation necessary to meet these ITER mission goals. The researchand development program includes activities directly related to specification of performancerequirements for ITER components as well as R&D required from other facilities, so as to provide asolid physics and technology basis for the efficient exploitation of ITER. It should be expected that,in its final form, the IRP will define the full range of R&D activities to be implemented in pursuit ofthe detailed scientific and technical goals. The emphasis in this (as in previous) versions of the IRP
is on the definition of the research program which will be implemented to satisfy the key scientificgoals, while the discussion of technology activities is limited to the Test Blanket Module (TBM)
testing program. While this latter element satisfies a key technical mission goal, future
developments of the IRP should extend the discussion of technology R&D activities to encompass
the wide-ranging research which will be undertaken on ITER tokamak, auxiliary and plant systems
in pursuit of the full scope of the project’s technical mission goals.
The IRP also aims to provide an analysis of the ITER facility’s capability to meet its mission andconsiders possible upgrade paths which may be required to address science and technology issues
supporting the achievement of specific mission goals. The current version discusses options forsuch upgrades and the potential impact of these upgrades on the project’s scientific research
program during the Operations Phase. In addition, a risk analysis has been carried out in which thelikelihood and consequences of negative R&D or exploitation outcomes are assessed, identifying,
where appropriate, possible risk mitigation and avoidance measures.

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The Research Plan is organized around a series of key thematic research areas, which are integrated
to form an exploitation plan for the facility. This exploitation plan has been embedded within theStaged Approach strategy developed during the revision of the ITER project schedule in 2015 –2016. The implications of this schedule strategy for the overall structure of the IRP and for thescheduling of the key activities within the research program, including the transition to operation
with deuterium-tritium fuel leading to significant fusion power production, are outlined in 1.3.
An attempt has been made to assign estimates of durations to the various elements of theexperimental program within the IRP, but these are necessarily uncertain and become more so as
the IRP progresses towards fusion power production, involving the study and optimization ofburning plasmas, an entirely new regime of laboratory plasma research to which ITER will provide
access. A perhaps more important output of the plan, at least at this stage of its development, is thelogical sequence developed for the programmatic activities and the identification of the necessary
prerequisites for the implementation of the major activities, in terms both of auxiliary hardware
availability and of output from previous steps in the plan.
As the project moves towards the Operations Phase, it will be essential to maintain a regular reviewof the ITER Research Plan, as results from the accompanying R&D program and, later, from theexperimental exploitation of ITER itself, become available. The IRP is thus to be viewed as a living
document that evolves with the project. In addition, the IRP serves as a baseline from which theconsequences of design or planning changes, or of unforeseen events, on the achievement of the
ITER mission can be determined.

1.2 Methodology
The initial version of the Research Plan (IRP-v1.0), presented to the ITER Council Science andTechnology Advisory Committee (STAC) in May 2008 [ITER_D_2FB8AC_v1.0, 2008], wasdeveloped during the 2007 - 2008 Design Review and was valuable in providing guidance to thecapabilities required to support the implementation of a scientific research program which could
address, and indeed fulfil, the project’s key scientific mission goals. However, the subsequentevolution of the project schedule, involving in particular, a change in strategy for the hardware
configuration at First Plasma and the installation of an all-tungsten divertor at the start of thenuclear phase, necessitated a significant revision of the IRP. Several updates were developed
adapted to the changes in project strategy and schedule and presented to STAC during 2009. IRPv2.1 was presented to STAC in October 2009 and, following minor modifications, was formallyincorporated into the ITER Baseline at the Extraordinary ITER Council in July 2010 as IRP-v2.2[ITER_D_2FB8AC_v2.2, 2009]. Although a further significant update (IRP-v2.3) was prepared andpresented to STAC in May 2010 which was fully adapted to the subsequently approved ITER
Project Schedule (IPS-2010), including a target date for First Plasma in late 2019
[ITER_D_346RPL, 2010], it did not complete the internal review procedure in time forincorporation into the ITER Baseline 2010, and so IRP-v2.2 has remained the reference Baseline
version of the Research Plan until the present document.
While key elements of the Research Plan have been revised to incorporate changes in the projectschedule and hardware configuration since 2010, perhaps most significantly the decision to initiate
the experimental program with an all-tungsten divertor, uncertainties over delivery dates for‘deferred’ components have prevented the development of a fully self-consistent revision of the
Research Plan since 2010. The complete revision of the ITER project schedule within theframework of the Staged Approach, undertaken in 2015 – 2016, has provided the basis for the23

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preparation of an updated IRP which is consistent with the expected ITER tokamak and facility
configuration during the period from First Plasma to full DT operation.
This revision of the IRP has been developed in collaboration with representatives of the Members’
fusion communities nominated by the Domestic Agencies (DAs) who contributed wide-ranging
expertise in fusion physics and tokamak operations to the analysis of the proposed experimental
program during the ITER Operations Phase. Three dedicated workshops were organized at ITER
Headquarters (HQ) involving ITER Organization Central Team (IO-CT) staff and the fusion
community experts during the period July 2016 to March 2017:Workshop 1: 26 – 28 July, 2016
Workshop 2: 14 – 17 February 2017
Workshop 3: 14 – 17 March 2017

These workshops allowed a detailed discussion among the IO-CT staff and fusion community
experts on the structure of the experimental program within the Research Plan, on the detailed
experimental activities to be implemented, on the associated risks and the preparatory R&D whichshould be carried out, and also provided an opportunity to review the estimates of experimental
time required to implement the foreseen program in the light of current operating experience on
existing fusion facilities.
The workshop participants were provided with a summary of the output from the Configuration
Workshop, held during 27 June – 1 July 2016, at which the facility hardware configuration during
the Staged Approach was developed by the IO-CT and IO-DA management and staff. In addition,
to provide background information on the key activities and issues within the existing ResearchPlan, the most recent report on the Research Plan, presented to STAC in 2014 [ITER_D_NSHUQ7,2014], was made available to the participants prior to the workshop. This report was used as a pointof reference for the analysis developed during the workshops. Complementary information on
experimental risks identified by earlier analyses of the Research Plan was also provided for review
by the Workshop participants [ITER_D_KTP2H5, 2013]. More detailed information on the
hardware configuration of the principal ancillary systems and on key aspects of ITER physics were
provided in a series of introductory presentations to the Workshop [ITER_D_TL5ACR, 2016]. To
facilitate the analysis of major activities within the Research Plan, the Workshop participants weredivided into several Working Groups each responsible for specific aspects of the experimental
program.
The Working Groups were requested to analyze the key activities required to develop the research
program following First Plasma, aligning the experimental activities with the planned hardware
configuration within the Staged Approach (see Figure 1.2-1) so as: (i) to commission all required
auxiliary systems with plasma to full performance (with pulse durations appropriate to the program
requirements); (ii) to develop the experimental capability required to achieve initial DT fusion
power production at levels of several hundred MW; and (iii) to propose a program of experiments
in DT plasmas aimed at meeting the project’s key mission goals in both inductive and non-inductive
plasma scenarios, incorporating also the needs of the TBM testing program. Workshop participants
were also requested to address:
 Whether requirements for system availability were consistent with the planned experimental
program; Estimates of the experimental time required for the various experimental activities within
the program;

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 Experimental or operational issues which would require further analysis within the ITER
R&D program to support the future development of the Research Plan; Risks to the Research Plan and possible mitigation measures.
The report presented here has been assembled from the analyses by the Working Groups of theelements outlined above.

Figure 1.2-1: Schematic of the structure for the first 10 years of ITER operations within the Staged
Approach.

The Research Plan presented in the current document addresses in detail the evolution of thescientific research program and of the TBM testing program during the 4 phases of the Staged
Approach, but also extends the analysis beyond the initial phase of DT operation to thedevelopment of long-pulse, high fusion power plasmas satisfying the project’s principal mission
goals.

1.3 Overall Structure of the ITER Research Plan Present Version
1.3.1

ITER Research and Operational Phases

As noted in 1.2, the Research Plan has been adapted to the Staged Approach strategy within therevised ITER schedule with the overall aim of exploiting the available experimental capability toallow the transition to nuclear operation in late 2035, following First Plasma in late 2025.
Furthermore, and as noted above, the development of the experimental program in DT plasmas isextended, on the basis of assumptions outlined in sections 2.6.4 and 2.6.5, to encompass thedevelopment of long-pulse plasmas at high fusion power and high fusion gain consistent with the
project’s mission goals.
Within the four stages of the revised operational schedule (including the longer-term extension of
the final stage to an extensive development program of DT fusion power production in long-pulse
DT scenarios and a research program in burning plasma studies), the principal goals are:
 First Plasma: achievement of the First Plasma milestone of plasma breakdown in hydrogenor helium with at least 100 kA of plasma current for at least 100 ms; during the subsequentEngineering Operation phase for commissioning of the Magnet systems to full current,
circular limiter plasmas of up to 1 MA might be attempted.
 Pre-Fusion Power Operation 1 (PFPO-1): extensive ‘commissioning with plasma’activities will be undertaken to establish diverted plasma operation in hydrogen/helium up toat least 7.5 MA/2.65 T, supported by a program of plasma control, diagnostics, ECRH andICRF commissioning (assuming that one antenna providing 10 MW of ICRF heating is25

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available), as well as the systematic development of a disruption management/ mitigation
capability to prepare for operation at higher currents; an option for operation at a toroidal fieldof 1.8 T has also been identified, which is under analysis and which could allow early accessto H-mode operation with up to 30 MW of installed ECRH + ICRF power; this approachassumes that dual-frequency operation (104/170 GHz) of up to 1/3 of the gyrotrons could be
implemented, or that up to 1/3 of the initial set of gyrotrons could be dedicated to low
frequency (110 GHz) operation (both solutions would imply associated Electron Cyclotron(EC) transmission and launching system modifications), if needed for plasma start-up and/or
heating.
 Pre-Fusion Power Operation 2 (PFPO-2): initial emphasis would be given tocommissioning of the Heating Neutral Beam (HNB), Diagnostic Neutral Beam (DNB) andICRF systems to full power, together with the necessary plasma control capability, to support
the subsequent program; development of H-mode operation in hydrogen, with helium as
back-up if this is not possible, reliable ELM control and divertor heat flux control would be amajor thrust during this period of the research program; a further key goal would be thedevelopment of high power L-mode operation to demonstrate the full technical capability ofthe device; systematic studies of plasma-wall interactions and issues such as plasma-facing
component erosion, redeposition, dust production, fuel retention and fuel removal would also
be pursued in preparation for the nuclear phase. If the option for operation at 1.8 T were
implemented, operation in H and in He H-modes would be possible at this stage, providing aphysics basis to compare H-modes in H and He, which could be very important if H H-modes
are not possible at higher fields in this phase. In addition, if H-mode operation at 1.8 T had
been achieved in PFPO-1, this would allow the evaluation of the effects of the 3-D fields
created by the TBMs in this phase, by comparing PFPO-1 (no TBMs) plasmas with those of
PFPO-2 (with TBMs), and the development of mitigation schemes for such effects, if
required.
 Fusion Power Operation (FPO): building on the experimental capabilities acquired during
the previous operational phase, the initial aim of this campaign would be to demonstrate
operation of high power H-modes in deuterium to prepare plasma scenarios for DT operation;
beginning with trace tritium experiments, a gradual transition to full DT operation would beundertaken as the fuel reprocessing throughput of the Tritium Plant increased; the goal of thefirst campaign of DT operation would be to demonstrate fusion power production of several
hundred MW for several tens of seconds at a Q value in the range of 5 – 10. In subsequent
experimental campaigns, the fusion gain would be optimized to achieve Q  10, the pulse
duration in inductive plasma scenarios would be extended towards the goal of 300 – 500 sburn required by the Project Specification, and hybrid and fully non-inductive scenarios
would be developed to achieve burn durations in the range 1000 – 3000 s with Q ~ 5.
The plan for each of the exploitation phases is based on a series of key research issues aroundwhich an experimental program has been developed to commission systems with plasma, todevelop certain control capabilities, to establish specific plasma scenarios, to address uncertainties
in the physics of plasmas at the ITER scale and in the burning plasma regime, and finally toperform detailed scientific studies of burning plasmas while optimizing their fusion performance.
There is greater uncertainty, of course, in relation to the details of later phases of the Research Plan,
since the results derived from earlier phases are expected to dominate the final planning ofsubsequent phases and, in addition, ITER is the first experiment which will provide access to thestudy of burning plasmas.

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In addition to the development of the scientific research program during the Operations Phase, thisdocument also presents, in 3.2, the research to be conducted within the Test Blanket Module (TBM)
Testing Program from the first operational tests of TBMs in PFPO-2 to the completion of the first
integrated tests of the operational behaviour of the blanket component for heat extraction andtritium breeding/ management, expected to take place during FPO campaign 3.
Section 4 of this report discusses a range of options for the implementation of upgrades to ITER
operational systems and indicates, where possible, how the experimental results from the Research
Plan can influence upgrade choices, as well as the implications of the upgrades for the subsequent
research program. In section 5, the IRP discusses the key research activities foreseen to proceed inparallel to the construction phase. These activities are intended to complete the design basis forcertain key auxiliary systems, to resolve issues impacting the efficiency of ITER operation, toprepare plasma scenarios to be used in ITER and to develop specific experimental techniques andtools required to achieve the foreseen scenarios. The discussion is focussed on key R&D issues,
which are grouped into several areas of fusion physics research. The identification of these key
R&D activities is closely linked to the resolution of physics and operational issues identified in thedevelopment of the Research Plan.
The present version of the IRP also presents several Appendices dealing with complementary
discussions of several issues closely related to the implementation of the research program.
Appendix A, Appendix C and Appendix D discuss specific issues associated with the plasma-wallinteraction: Appendix A discusses the use of deuterium ‘spiking’ of hydrogen plasmas during
PFPO-2 to provide an improved quantitative estimate of fuel retention, while Appendix C presents
the considerations relating to the management of heat loads on the ITER divertor and Appendix D
describes the establishment of a routine ‘reference pulse’ which would be of particular significancefor characterizing the evolution of the condition of plasma-facing materials. Appendix B reviewsthe experimental implications of our understanding of the scaling of the H-mode power threshold to
ITER and of the impact of TF ripple on H-mode performance, while the physics considerations
related to H-mode operation at 1.8 T, an option for the research program which has emerged
recently, are presented in Appendix F.
In view of the proposal within the revised ITER schedule to eliminate the option for a later
inclusion of an LHCD system in ITER, Appendix E presents a brief review of results of modellingstudies of long-pulse and non-inductive DT scenarios in ITER. In particular, the predicted plasma
performance with and without inclusion of LHCD in the H&CD upgrade options are compared interms of the Q-value achieved and of the proximity to the fully non-inductive state. An extensive
discussion of the plans for the phased installation of the baseline H&CD systems, of options formodifications to the agreed baseline installation schedule and capabilities, and of the schedule
implications of various upgrades to the baseline systems is presented in Appendix G. The strategy
developed within the revised schedule for the phased installation of the baseline Diagnostic
capability is presented in detail in Appendix H. A brief summary of considerations related to theavailability of tritium from external sources to support the planned DT experimental program
during the period from the late 2030’s to the early 2050’s is given in Appendix I. Appendix J
provides an updated version of the Research Plan Risk Register, which deals with the mostsignificant risks to the successful implementation of the planned experimental program and, inparticular, to the timely achievement of significant fusion power production.
1.3.2

Assumptions on the Phasing of ITER Operational Capabilities

The central element of the revised ITER schedule within the Staged Approach is the phased
installation of the operational capability of the ITER tokamak and of the ITER facility. A detailed27

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discussion of the Plant Configuration proposed for each phase of the Staged Approach waspresented to the STAC in 2016, and the baseline ‘Plant Configuration’ within the revised schedule
has been formalized in [ITER_D_TVG7YK, 2017]. Figure 1.3-1 provides a schematic overview ofthe evolution of the plant configuration during the Staged Approach, indicating in yellow options
for PFPO-1 which are not yet in the baseline, but are currently under evaluation. The major
elements impacting on the implementation of the Research Plan are summarized here for ease ofreference. Further details relating to the phasing of the H&CD systems and to the sequence ofDiagnostic installation are presented in Appendices Appendix G and Appendix H respectively. The
following evolution of the auxiliary systems has been agreed to support the experimental program
discussed in sections 2.4, 2.5 and 2.6:
(I) First Plasma (H/He) (FP):









Initial set of PF/CS converters
First Plasma Protection Components (FPPC) to provide required protection of in-vessel
systems and vacuum vessel
Central Control System fully operational
H&CD installed power: PECRH = 8 MW from 1 upper launcher (6.7 MW coupled to theplasma)A limited subset of the baseline Diagnostic capability, essentially to characterize plasma
breakdown and to provide limited plasma control and machine protection
Fuelling: gas injection capability
Vacuum: 6 cryopumps incorporating activated charcoal
Vacuum Vessel baking and preliminary Glow Discharge Cleaning system(II) Pre-Fusion Power Operation (H/He) 1 (PFPO-1):











PF/CS converters upgraded to full performance
Shielding blanket/beryllium first wall and all-tungsten divertor
H&CD installed power: PECRH = 20 MW (all launchers)
A subset of the baseline Diagnostic capability, with the emphasis on plasma control andmachine protection
Fuelling: full gas injection capability and at least 2 pellet injectors
Disruption mitigation system: hardware commissioned
Error field correction coils: hardware commissioned
Vertical stabilization coils: hardware commissioned
In-vessel viewing system; hardware commissioned
An appropriate Be handling capability will be provided

All of the listed systems will be required to support the experimental program. In addition, thefollowing options are being studied, with the aim of implementing these through later Project
Change Requests (PCR) if their feasibility and benefit to the research program is confirmed:

Up to 1/3 of the installed ECRH power to have a low frequency capability (104 or 110
GHz) either via a set of 8 dual-frequency gyrotrons (104 GHz), or a set of 8 dedicated low
frequency gyrotrons (110 GHz), if needed for plasma start-up and/or heating.
Acceleration of one ICRF antenna to provide up to 10 MW injected power capability at
40 – 55 MHz.

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

A subset of 9 of the ELM coil power supplies to allow initial ELM control studies and an
improved error field correction capability.

The experimental program detailed in 2.5.4 assumes that these additional capabilities will beavailable.

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Figure 1.3-1: Schematic of the ITER ‘Plant Configuration’ at each phase of the Staged Approach, with thecolour code indicating the stage at which each system becomes fully operational. (*) refers to temporaryconnections to delivery and exhaust of non-tritiated gases during PFPO phases.

(III) Pre-Fusion Power Operation (H/He) 2 (PFPO-2):








H&CD installed power: PNB = 33 MW, PECRH = 20 MW, PICRF = 20 MW (two antennas)
Diagnostic Neutral Beam (DNB) system
An extended set of Diagnostics for physics studies
Fuelling: 2 additional pellet injectors
ELM control coil system: hardware commissioned
First set of TBM modules (EM-TBM) installed for initial phase of testing program
Tritium Plant undergoing integrated non-active commissioning(IV) Fusion Power Operation (D/DT) (FPO):
It should be expected that essentially all elements of the tokamak and its auxiliary systems will havebeen brought into operation during the PFPO phase in order to allow the efficient implementation of
the D and DT phases of the program. Significant new capabilities that would be available for the
nuclear phase include the measurement capability for 14 MeV neutrons; since this aspect of theneutron calibration will be scheduled just prior to the start of the nuclear phase of operations, and arange of fusion product diagnostics (see Appendix H). A principal goal of the PFPO phase is toensure that all of the H&CD capability is commissioned to full power to at least ~ 50 s before thenuclear phase and that all elements of the fuelling systems are operational with all required gases
except deuterium and tritium. As will be discussed in sections 2.4 and 2.5, key aims of the PFPO
phase include commissioning of all aspects of magnetic control, several aspects of kinetic control(e.g. control of divertor heat exhaust), demonstration of reliable and effective disruption mitigation(including runaway electron mitigation), and demonstration of ELM suppression/ control. Thus,
essentially all of the plant and ancillary systems should be in an advanced state of operation by the
beginning of the nuclear phase. The major plant commissioning activity that would proceed inparallel with nuclear operation would be the expansion of the processing throughput of the Tritium
Plant. This would rise gradually during the deuterium experiments and into the initial period of DT
operation. In addition, if essential upgrades have been identified during PFPO to achieve the Q = 10
goal, it is likely that these will be implemented either before or in the early stages of FPO; this may
require specific commissioning time to be dedicated to them.
1.3.3

Key Research Issues

1.3.3.1 Key Research Issues during Operation
The Research Plan for the exploitation phases is developed around a sequence of ‘commissioning
with plasma’ activities and a series of key research issues. Some of these issues will be addressed inparallel so as to use experimental time efficiently, but many must be resolved before proceeding toaddress subsequent issues.
In the PFPO, the first physics objective will be to develop plasma scenarios suitable for systemcommissioning, particularly of the heating and current drive systems. This will necessarily involve
extensive commissioning of the plasma control and protection systems (Central Interlock System,
CIS; Central Safety System, CSS, Plasma Control System PCS) to ensure safe and reliable
operation of the tokamak and investment protection. Such commissioning includes the use of the invessel coils for vertical stabilization of highly shaped plasmas and studies of the level of error field30

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in ITER, together with the determination of the impact of error fields on ITER plasmas, and of theuse of the dedicated coil set to correct them routinely. In parallel with these commissioning
activities, the characterization of disruptions, the development of reliable disruption prediction, andthe commissioning and optimization of the disruption mitigation system (DMS) will be an essential
research activity accompanying the gradual increase in plasma parameters during this period.
Commissioning of diagnostic systems and data validation will be important early activities toprovide the measurement capability for plasma control, disruption detection and more generalmachine protection: for example commissioning of a protection system for the all-metal plasmafacing components will be important in advance of the introduction of significant heating powerand higher plasma thermal energies.
Once an adequate measurement, control and protection capability is in place, commissioning of the
H&CD systems will be a priority to provide access to higher plasma parameters and, eventually,
higher confinement. Initial physics studies carried out in this phase will also be the first opportunity
to study tokamak plasma behaviour at the ITER scale. While there is still considerable uncertainty
over the scaling of the H-mode power threshold to ITER (see, e.g. Appendix B) the experimental
data indicates that it could be possible to obtain H-modes in helium plasmas with ECRH heating
only. The option to operate ITER at 1.8 T, with an adapted H&CD capability (i.e. the possible
additional 10 MW of ICRF) could open a path to early studies of hydrogenic H-modes, notpreviously considered possible during this early operational period. H-mode access would also
permit an early assessment of crucial issues, such as ELM control, while potentially allowing abetter definition of the requirements for future heating and current drive upgrades. As discussed in
2.2, there are many important plasma-wall interaction issues which will be of interest, even within
the limited range of plasma parameters accessible during this period.
The additional H&CD and Diagnostics capabilities to be installed in advance of PFPO-2 willconsiderably expand the range of plasma scenarios and plasma parameters available for study, andwill enrich the capability for physics studies at the ITER scale. ‘Commissioning with plasma’activities will continue during this period, since it will be important to commission as many of thetokamak and plant systems as possible and test them thoroughly in advance of the transition to
D/DT operation. Thus commissioning of plasma control, disruption management, machine
protection, H&CD and Diagnostic systems will continue to feature extensively, but moresophisticated aspects of plasma control will be explored, including NTM suppression, sawtooth
control, and possibly physics issues related to RWM control, and certainly divertor heat flux
control. A more extensive program of H-mode studies and, in particular, of ELM control will be
implemented, with the emphasis on studies of hydrogenic H-modes if access to H-mode inhydrogen plasmas proves possible. Extension of the plasma operating range to high current and tolong pulses (at lower currents) will be key aims of the program during this experimental campaign:
first operation at 15 MA/5.3 T should be demonstrated in L-mode, while the current drive
capabilities of the HNB systems could allow pulse lengths of several hundreds of seconds in bothL- and H-modes, if sufficient time for commissioning the H&CD systems to long-pulse operation isavailable. In addition, initial studies of current drive efficiency and formation of target q-profiles forhybrid and/or non-inductive scenarios could be launched during this phase in preparation for thedevelopment of these scenarios in DT plasmas. PWI studies will be of particular significance during
PFPO-2, as quantitative studies of power deposition and of processes such as erosion, material
migration, fuel retention (and removal) and dust generation are expanded with a view to developing
a quantitative understanding of these issues in advance of DT operation. A wide-ranging program of
PWI studies with a dedicated experimental program is therefore anticipated.

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Following the transition to the FPO phase, plasma operation will initially focus on experiments indeuterium: the deuterium phase is likely to prove the first opportunity to perform detailed
development of H-mode scenarios for later use in DT. Nevertheless, since the key goal of this phaseis to transition to 50:50 DT plasmas and significant fusion production as rapidly as possible, theduration of the deuterium program will depend strongly on the pace of commissioning of the
Tritium Plant, which is likely to begin the experimental campaign with a very limited throughput
that will then grow over time. Therefore, while physics studies in deuterium L- and H-modes will
provide extensive opportunities for the exploration of plasma behaviour at the ITER scale, once anadequate supply of tritium becomes available, it is favourable from the programmatic and mission
perspectives to launch the development of plasma scenarios with increasing tritium fractions,though initial experiments with trace tritium in deuterium plasmas can provide a valuable tool forthe investigation of physics processes governing tritium fuelling and transport. Alternatively, if thetritium throughput of the T-Plant rises more slowly than required, additional time could be allocatedto scenario development in deuterium with the aim of minimizing development time required in
DT. During this program, control of ELMs, core MHD and (simulated) plasma burn will beestablished and combined into fully integrated scenarios.
This phase will initially focus the experimental R&D on conventional inductive scenarios but also
includes the initial development of advanced and hybrid scenarios. The extent to which the latterdevelopment is carried out will depend on the available experimental operational time until asignificant tritium throughput is available. The aim of the present ITER Research Plan is to proceed
to full DT operation as rapidly as allowed by progress in commissioning of required systems and in
the experimental program. Thus priority is given to establishing an adequate basis for thedemonstration of significant fusion power production and fusion gain for several tens of seconds inthe most promising plasma scenario (currently assumed to be the inductive ELMy H-mode) and theextensive development of more advanced scenarios is deferred until later in the DT phase.
Once an adequate tritium throughput is established, the transition from deuterium towards full DT
plasmas will follow a route of gradually increasing both plasma current and tritium fraction,
allowing time for issues such as power handling, burn control, fuel mixture control and helium
exhaust to be integrated into scenarios established initially in the deuterium phase. The firstprogrammatic goal is to develop an integrated scenario capable of producing several hundred MWof fusion power for several tens of seconds. Achieving this goal during the first FPO campaign willbe the striking demonstration of ITER and magnetic fusion capabilities and potential.
Once this level of fusion performance is achieved, the focus of the program will be on exploring
optimization of DT scenarios, aiming to achieve the main mission goal of demonstrating Q ≥ 10,
which might firstly be accomplished in relatively short pulses of several 10s of seconds. In parallela detailed study of burning plasma behaviour in the most performing scenario would be undertaken,
the first steps into a new regime of fusion plasma studies. Thereafter, the development of plasma
performance and the exploration of burning plasma physics would be intertwined in a program
designed to optimize Q and extend the pulse duration towards several hundred seconds.
Complementing the development of long-pulse burning plasma scenarios, experimental time would
be devoted to the development of hybrid (or improved H-mode) scenarios and fully non-inductive
scenarios. Achieving both long duration burns and a Q-value approaching 5 in the latter scenario islikely to involve a program of at least several years experimentation, while the hybrid mode may
offer a quicker route to long-pulse operation if the physics observed in current experiments
extrapolates sufficiently well to the ITER scale and to the burning plasma environment.

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An important element of the Research Plan during DT operation is, of course, the study of a wide
range of burning plasma physics issues:








Waves and energetic particle physics;
Fuel mixture control and helium exhaust;
Self-heating and thermal stability;
Macroscopic stability physics and control;
Multi-scale transport physics;
Physics of the plasma-boundary interface;
Integrated burning plasma scenario development.

Additional key physics issues for advanced modes include current profile control and MHD control
at pressures above the no-wall β-limit. These studies will not only address issues related to the
ITER core mission but also leverage the unique capabilities that ITER provides within theworldwide fusion program. Examples of such capabilities include operation at low normalized
gyroradius, operation with a low collisionality pedestal coupled with a high-density divertor, andoperation with a large population of isotropic energetic particles.
These detailed physics studies will be complemented by a comprehensive technology program, inparticular on the tritium breeding Test Blanket Modules, which are the subject of an extensive
research program in the ITER Members and which will be a further significant contribution by the
ITER Project to the development of fusion energy.
1.3.3.2 Key Research Issues during Construction
The key issues identified for the supporting Physics research program accompanying ITER
construction have been grouped into four areas: H-mode issues; Plasma-wall interaction issues;
MHD instability mitigation/control issues; and Scenario development issues. It is recognized thatthe four types of issues are linked. Indeed, one of the overall goals of the program is to develop theintegrated modelling capability for ITER, within the Integrated Modelling and Analysis Suite(IMAS), to provide a ‘flight simulator’ that can be used as part of the preparation of experiments
and that can be validated by the results in present fusion facilities, in advance of its application forthe preparation of ITER plasma discharges. In addition, the R&D activities required to support thedevelopment and construction of the Test Blanket Modules are reviewed (see 3.2.1.2).
The key research issues during construction are discussed briefly in section 5 of the Research Plan.
Resolution of these might have an impact on ITER in at least one of three ways: affecting thedesign of baseline components of the machine – though with the progress in the detailed design of
ITER systems over the past 10 years and the transition to construction, the scope for influencing thebaseline design is much reduced; influencing the choice or design of upgradeable components; orleading to improvements in the plasma scenarios and experimental plan foreseen for ITER, thereby
improving the efficiency of ITER operation. Issues in the first category are obviously time-critical their resolution must rely either on already existing capabilities or on modest extensions of thosecapabilities. The second category also has time constraints, but these are more difficult to define
precisely: decisions on upgrades generally require long lead times that are a function of theircomplexity (i.e. for design, fabrication, and installation), and the supporting R&D may require
additional time for new hardware as well as analysis. The last category encompasses a range ofchallenging issues whose outcomes impact the experimental time allocated in the ITER schedule to33

ITR-18-003

the development of plasma scenarios which encompass adequate solutions at the ITER scale. Their
resolution could also indicate that other upgrades need to be considered.
The list of key issues presented in the Research Plan is clearly not exhaustive. There are many otheruncertainties in the physics and technology basis for ITER that can also impact performance andwhich are being addressed in the extensive research program being carried out in parallel to ITER
construction. The items listed here, and discussed in more detail in section 5, are those currently
considered to be the highest priority, in particular with regard to their perceived probability ofleading to the need to modify or upgrade the facility.

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2 Research Program during Operations
2.1 Overall Structure of the Experimental Program during Operations
The basic structure of the experimental program within the Research Plan has been maintained
under the adaptation to the revised schedule within the Staged Approach. The program begins with
3 experimental phases (including First Plasma) in hydrogen/helium (i.e. non-activating fuels) inwhich all of the tokamak, plant and auxiliary systems are commissioned with plasma to highperformance (not all aspects of system performance can be commissioned to the highest level inadvance of fusion power production). Following the transition to the use of D/DT fuels, an initialcampaign is designed to achieve significant fusion power production as rapidly as possible, and thisis followed by a sequence of experimental campaigns on a two-yearly cycle. In these campaigns, inwhich the ITER scientific and technical mission goals are pursued (including the TBM testing
program), burning plasma physics is studied in detail, and the physics basis for Demo design andoperation is developed. The present Plan develops the exploration of the DT phase within theachievement of the principal scientific mission goals and so limits the time horizon of the program
to the first 3 experimental campaigns in DT plasmas, as illustrated in Figure 2.2-1. The broad aimsof each phase of this experimental program have been outlined in section 1.3.1, and these are
expanded in some detail in sections 2.5.4.1, 2.5.5.1 and 2.6.1.
The structure on which the Staged Approach converged as a result of the schedule revision
procedure implemented by the ITER project management during 2015 and 2016 is illustrated in
Figure 1.2-1. This was developed on the basis of guidance from the Members on their anticipatedbudget constraints during the Construction Phase and the implications for the resources expected to
be available to IO-CT and IO-DAs, but also on the understanding of ITER plasma operation andexperimental R&D developed over the past 10 years. This understanding has been developed
through the elaboration of previous versions of the Research Plan and additional experience gained
through the fusion physics R&D programs in the Members’ fusion communities, which haveprovided insight on the experimental time that will probably be required to progress from First
Plasma to a level of plasma performance and operational efficiency in ITER which would: (i)confirm that the tokamak systems were performing satisfactorily, (ii) provide confidence that an
adequate level of plasma performance had been achieved in ITER, and (iii) that the necessary levelof experimental expertise (integration and control of plasma scenarios) had been demonstrated tojustify the transition to DT operation and fusion power production. The resultant schedule towards
DT operation incorporates both the project level and scientific/ technical considerations on the path
towards DT operation.
The resultant structure of the experimental program, illustrated in Figure 2.2-1, therefore consists of:







First Plasma Campaign: ~1 month, plus 6 months of Engineering Operation including,possibly, some plasma operation
Pre-Fusion Power Operation 1: 18 months of hydrogen/helium plasmas
Pre-Fusion Power Operation 2: 21 months of hydrogen/helium plasmas
Fusion Power Operation 1: 16 months of initially D and subsequently DT plasmas
Fusion Power Operation 2: 16 months of DT plasmas
Fusion Power Operation 3: 16 months of DT plasmas

Beyond this period, which is the subject of the analysis presented in succeeding sections, the ITER
experimental program would, of course, be expected to implement extensive scientific and technical35

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R&D in pursuit of its overall mission goal to demonstrate the scientific and technological feasibilityof fusion energy for peaceful purposes and to provide key information for the further developmentof Demo. This longer-term development of the ITER program is beyond the scope of this version of
the Research Plan.
2.1.1

Integration of the Research Plan into the Staged ApproachAn essential element in the integration of the ITER Research Plan into the Staged Approach is thelink between the hardware capabilities of the ITER facility (the ‘Plant Configuration’) at each stageof the Operation phase and the corresponding objectives. Once again, a balance has been struck
between the project-level constraints on the procurement schedule and the operational requirements
to fulfil overall objectives for the PFPO phase which would support a timely transition to DT
operation, with a target date for this transition of late 2035. The evolution of the ITER facility
hardware configuration has been outlined in section 1.3.2 and, as noted there, formalized in[ITER_D_TVG7YK, 2017], which defines the baseline configuration of the ITER facility at eachphase of the Staged Approach. In addition, several options are being developed within the baseline
configuration that are expected to be formally adopted through the project-level Project Change
Procedure once their technical benefits and the technical/management feasibility of theirimplementation are confirmed. The key elements of these options, which influence the developmentof the Research Plan, and which have, therefore, been integrated into the analysis of the Plan (i.e.
assumed to be available to the experimental program) are summarized in section 1.3.2.
In addition to the logical structure defining the steps necessary to progress from First Plasma to full
performance in DT plasmas, the Research Plan makes estimates of the duration of each of the steps
within the experimental program to obtain an overall estimate of the operational time required forsuccessive phases of the program. This analysis is clearly subject to substantial uncertainties so far
in advance of ITER exploitation. Nevertheless, in discussion with the fusion community experts
participating in the Research Plan Workshops, a first estimate of the experimental time required by
the principal activities was assembled, which forms the basis of the estimates given in the course of
the discussions in later sub-sections of section 2.
As a point of reference, an estimate of the maximum number of experimental days available in
PFPO-1, PFPO-2 and FPO is given in Table 2-1. The basis for this estimate is that applied in earlier
analyses of the Research Plan, derived from the assumptions developed in the ITER RAMI analysis[ITER_D_28WBXD, 2012]: in 2-shift operation, which forms the basis of operational planningwithin the Staged Approach, it is assumed that the operational pattern is 12 days of experimental
operation and 2 days of short-term maintenance within a 14-day cycle.
Table 2-1 – Overview of experimental days within the Research Plan campaigns
Campaign
First Plasma Campaign (nominal 1 month)
Pre-Fusion Power Operation 1 (18 months)
Pre-Fusion Power Operation 2 (21 months)
Fusion Power Operation 1 (nominal 16 months)
Fusion Power Operation 2 (nominal 16 months)
Fusion Power Operation 3 (nominal 16 months)

Days
25
470
545
415
415
415
2285

Total

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Within a 2-shift operational day, a 30-minute pulse repetition time, for burn (or flat-top for nonactive operation) durations of less than 450 s allows, in principle, 32 plasma pulses. When corrected
for the 60% ‘Inherent Availability’ assumed for the RAMI analysis, and the loss of time associatedwith disruptions/disruption recovery, a total of 13 ‘good’ plasma pulses per day results. Later in the
FPO phase, as pulses with burn durations beyond 450 s are developed, the number of pulses per day
will, of course, be reduced if a 25 % duty cycle is maintained. However, this reduction in ‘good’
pulses per day achieved within the constraint of a 25% duty cycle may be compensated, to someextent, by a reduction in the frequency of disruptions; this will necessarily be substantially lower in
the FPO phase than in the PFPO phase as a result of the progress made in disruption management,
as outlined in later sections, and of the need to avoid significant loss of experimental time due to theimpact on in-vessel components of disruptions at high fusion power.
The estimates given in Table 2-1 are, of course, likely to be an upper bound to the number of days
actually available to the experimental program, since the system performance assumed in the RAMI
analysis relates to the operation of mature systems, but they nevertheless provide a basis forcomparison with the estimates of the experimental time required for the main activities within the
Research Plan presented in the following tables.
The transition from the PFPO phase, during which non-activating fuels are used, to the FPO phase,where deuterium and deuterium-tritium plasmas form the basis of the experimental program, is, ofcourse, a key transition in the Research Plan. Within the Staged Approach, this is likely to occur
following Integrated Commissioning IV, in the latter half of 2035. In preparation for supporting theexperimental program following this transition, it is planned that the Tritium Plant will take thesignificant step of embarking on the commissioning of its operation with tritium in spring 2033 (seeFigure 2.2-1). It is expected that early in FPO campaign 1 (2036), the T-Plant will be able to supply
trace tritium amounts for the ITER experimental program. In the course of 2036, a fraction of the
Isotope Separation System (ISS) will be operational and will be able to provide pure tritium for
ITER operation, reaching a daily fuel throughput of the order of 10% of the final throughput during
the FPO campaign 1 (the final throughput is specified as 200 Pam3s-1 (~ 1023 D/T atoms s-1) for a
300 – 500 s Q = 10 plasma with a duty cycle of 25%). The initial throughput will be sufficient toallow an expanding program of DT experiments to be developed during the latter half of the first
FPO campaign, leading to significant fusion power production of several hundred megawatts during
the campaign. This framework for the commissioning and operational start-up of the Tritium Plant
forms the basis for the development of the FPO program presented in section 2.6.

2.2 Plasma-Wall Interaction (PWI) Issues for the Research Plan
2.2.1

Characteristics of Helium Operation

The effect of helium interactions with plasma-facing metals has received considerable attention inrecent years due to the strong trapping exhibited by He in most metals and its tendency to cluster
and form bubbles, which leads to strong morphology changes (Figure 2.2-2). The possible
formation of a porous sub-surface structure (populated by a large fraction of voids/bubbles) and theobservation of a strong reduction in the thermal conductivity of the surface in such cases [Cui,2017] has raised concerns about possible reduction of the transient damage thresholds (especiallyfor ELMs) after extended He plasma operations, as considered during the PFPO phases. Heliuminduced embrittlement has indeed been observed after high-flux plasma exposure in linear devices
showing that He plasma exposure of W leads to a reduced threshold for surface cracking.

37

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Figure 2.2-1: Operations Plan within the Staged Approach.

38

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However, ELM-simulation experiments in electron-beam facilities have shown that for ITERrelevant high cycle number (~106), the cracking threshold of tungsten in terms of energy density isin any case very low [Wirtz, 2017]. This indicates that surface cracking might appear sooner if
extensive He operations were performed, but that it would in any case be likely to occur whether ornot Helium operation is performed, as He will be produced during DT operation.

Figure 2.2-2: Illustration of the range of He-induced morphology changes on tungsten as a function oftemperature. A dense network of nanobubbles is usually formed at temperatures below 600-700 K (left:
image from [Miyamoto, 2011]), fuzz formation proceeds in the range 900-1900 K (centre: images from[De Temmerman, 2012]), large voids are typically formed at higher temperatures (right: image from [DeTemmerman, 2012]).

The effects of He exposure on the mechanical properties of W have been studied using nanoindentation [Yoshida, 2005; Bernard, 2015] and revealed measureable increases in the hardness far
from the implantation depth of the incident He particles. To date, however, the existing data is stillinsufficient to build a coherent picture of the effect of prolonged He (and plasma in general)
exposure on the thermo-mechanical properties of W and how this might affect the power handlingcapabilities of the material in the long-term. It should be mentioned here that He effects on metals
are strongly temperature-dependent. It is expected that the divertor heat fluxes, and hence surface
temperatures, during pure He operations, as presently foreseen during PFPO-1 and 2, will be
substantially lower than during high power DT operations where the He from fusion reactions will
be present in the divertor exhaust. Any knowledge obtained on ITER regarding He-W interactions
during non-active operations will therefore be difficult to extrapolate to burning plasma conditions.
A dedicated He campaign was performed in ASDEX-Upgrade in 2015 in part to study plasma-wallinteractions in a full metal machine during He operations [Hakola, 2017]. A series of samples was
installed on the ASDEX-Upgrade outer divertor target manipulator and exposed at the outer strike
point location. The most noticeable result of this campaign was the observation that the outer strike
point region was a deposition-dominated area during He operations, in sharp contrast to its status as
an area of erosion during similar D discharges. More precisely, thick boron coatings (boronization
is used for wall conditioning on ASDEX-Upgrade) were observed on the samples, implying an
increase in wall erosion during He operation. The most likely cause for this observation is thehigher average charge state of He ions impacting main wall surfaces (and higher mass) compared39

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with D, leading to enhanced boron sputter erosion of structures around the ICRF antenna and/or
enhanced ELM-induced erosion of boron coatings and thus to the outer divertor becoming a net
boron deposition zone for He plasmas. A confirmation of these results in JET would be very
valuable. Translated to ITER, these results indicate that the material migration pattern during pureHe operations could be very different from that expected during DT operations casting doubt on theusefulness of He operations for studies of beryllium co-deposition. In addition, given the differencein the retention mechanisms of H isotopes and He in plasma-facing materials, limited information isgained during He operations regarding fuel retention.
It may also be noted that divertor detachment physics in He plasmas differs significantly from thatin hydrogenic plasmas [Loarte, 2000; Pitts, 2003; Wischmeier, 2003]. As a result of the higher
ionization energy of He, the presence of a second ionization stage, and the lack of molecularchemistry, the loss processes which dominate plasma dissipation leading to detachment in a He
plasma are distinct from those in hydrogenic fuel mixes. In hydrogen, plasma detachment proceeds
first by means of momentum losses due to CX processes, whilst in helium, the main mechanism isradiative loss from electron-impact excitation and ionization events. Ionization mean-free paths are
longer in He in comparison with hydrogenic plasma so that He neutrals may more readily escapefrom the target vicinity, radiating along the separatrix and above the X-point area, starving regions
downstream of power. The net result is that the dynamics of plasma detachment are likely to be
sufficiently different that power load control techniques developed in He plasmas may have limited
applicability to operation at high power in D or DT and, thus, that H H-mode plasmas are preferable
to develop such control techniques in this phase. On the other hand, the additional divertor radiation
obtained in He means that operation is possible at higher input power before divertor detachment
control is required (Appendix C) thus facilitating high power operation, if issues are identified in
the control of power loads for H H-mode plasmas that require additional systems at this stage.
2.2.2

Steady-state near and far Scrape-Off Layer (SOL) power widths (1/Ip scaling)

The radial profile of parallel power flow, and the extent to which main wall interactions will be
important in ITER are still active areas of research throughout the fusion community and consensus
has not yet been obtained. A particular concern is the very narrow near-SOL inter-ELM heat flux
width (q,near) now being predicted for ITER in the baseline burning plasma scenario (Ip = 15 MA)
from a new scaling based on high resolution IR thermography measurements at the outer divertor
target of several tokamaks [Eich, 2013]. As discussed in Appendix C, these new observations,unavailable at the time of writing of the previous ITER Research Plan, now extrapolate to q,near ~1.0 mm, a factor of ~3 lower than assumed in all plasma boundary modelling used to scope theoperational window of the ITER divertor. The scaling, in which q,near  1/Ip with little or nodependence on any other key SOL parameter previously found in similar scaling attempts (e.g.
PSOL, separatrix density, R) is strongly supported by analytic modelling based on neoclassicalmagnetic drift theory [Goldston, 2012]. However, very recent gyro-kinetic simulations at the ITER
scale find a strong major radius dependence, giving q,near ~ 5-6 mm in 15 MA H-modes [Chang,
2017].
A simulation study [Kukushkin, 2013] using the SOLPS plasma boundary code of the consequencesof a very small q,near showed, as expected, that the upstream density would have to risesignificantly if extra dissipation is required in the divertor to reduce the peak heat flux density. This
increase may be beyond what is tolerable from the operational density limit point of view or fromthe standpoint of maintaining adequate confinement. Moreover, the addition of divertor monoblock
shaping in the high heat flux regions of the target exacerbates the situation and will require still
higher dissipation if target heat flux densities are to remain below the 10 MWm-2 technology limit
40

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for steady state power handing. The research community, recognizing the importance of this issue
for ITER, is actively engaged in finding consensus as to what should be expected in ITER at highperformance regarding q,near. Further plasma boundary simulations, beyond those in [Kukushkin,2013] are also required to study the extent to which impurity seeding could be used to provide theextra dissipation in the event of very narrow heat flux channels.
Ultimately, ITER operation will be required for the real q,near to be properly characterized. Highspatial resolution IR diagnostics viewing a small toroidal region of the divertor vertical target areforeseen and will be operational for PFPO-1. Recent observations on ASDEX-Upgrade [Sieglin,
2016] have shown a similar parametric dependence for q,near in L-mode discharges to that
previously found in H-mode, but with about twice the absolute magnitude. This is reassuring fromthe point of view of thermal load specifications, as this assumption was adopted for ITER.
Moreover, the new measurements find a dependence of q,near on the edge plasma temperature,
which is speculated as one explanation for the higher values in L-mode in comparison with Hmode. The implication is that measurements of divertor power widths in ITER L-modes can be usedas guidance for the values to expect in H-mode. The ASDEX-Upgrade data also demonstrate astrong dependence of the divertor heat flux spreading parameter, S (due to perpendicular heattransport in the divertor) on divertor Te and ne and show that this parameter is a useful tool tocharacterize target heat loads.
A closely related issue is that of the SOL power width and divertor target wetting in the presence ofmagnetic perturbations for ELM control/mitigation. Dimensioning of the divertor targets andperformance has been almost exclusively performed using the SOLPS code in 2-D and thus takes
no account of 3-D fields. Some analysis for ITER has been performed [Schmitz, 2016], using the
3-D code EMC3-Eirene, showing how divertor fluxes are redirected into a 3-D pattern of helicalmagnetic footprints at the targets. At maximum perturbation strengths (maximum ELM control coilcurrents), these ‘fingers’ can extend well beyond the axisymmetric heat flux decay profile, evenwith plasma screening accounted for. These studies, however, have been extended to date only up
to the high recycling regime. Further developments in EMC3-Eirene are required to access truly
dissipative regimes (e.g. including strong seeding and recombination). Moreover, understanding of
how the divertor heat flux patterns in the presence of 3-D fields are affected in highly dissipative
regimes is hampered by the fact that ELM suppressed regimes on current devices are found to date
only at low collisionality. Experiments in which higher density divertor plasmas have been run in
the presence of 3-D fields generally find that the heat flux perturbation at the target introduced by
the magnetic perturbations is removed in-between mitigated ELMs, when the plasma detaches andthe profiles are very similar to those found in the absence of the perturbation. However, it isexpected that the removal of ELMs during complete suppression, supposing such conditions can befound on current devices, will strongly modify the upstream outward particle and heat fluxes so thatthe divertor target profiles which would be found in this case are presently unknown.
The current status of this research was summarized by the ITER Organization in a report to STAC
in 2016 and this issue has been given high priority in the International Tokamak Physics Activity(ITPA) Divertor and SOL and Pedestal Topical Groups, with joint experiments and modelling
expected to provide substantially improved understanding in the coming years.
Concerning the far-SOL power and particle fluxes on ITER (expected to be dominated byfilamentary, or blobby cross-field convective transport), this remains an area of uncertainty and isbeing actively studied, notably within the ITPA Divertor and SOL Topical Group. It has important
implications for the magnitude of wall heat and particle fluxes and hence on power handling,material erosion/migration and ICRF coupling. Recent multi-machine efforts have identified a link
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between the appearance of upstream density shoulders and the degree of divertor plasma
collisionality in the L-mode [Carralero, 2015]. Similar behaviour has also been found inter-ELM in
H-mode plasmas, but the results are less conclusive [Carralero, 2017], with divertor collisionality
seemingly a necessary, but not sufficient condition for upstream density shoulder formation. No
data are available for the case of perturbed magnetic fields for ELM control; measurements are
difficult and rare under H-mode conditions.
Considerable progress has been made, however, in demonstrating the universal nature of far SOL
turbulence in tokamaks and a model has been developed which successfully describes thesestatistics [Garcia, 2016]. It is also the case that SOL turbulence simulations are increasing
considerably in sophistication, with several 3-D codes being developed capable of including
sheared (X-point) geometries. It is likely that continuing R&D, both experimentally andtheoretically, will lead to substantial improvements in the understanding of main chamber fluxes by
the time ITER operation begins.
2.2.3

Edge localized mode (ELM) power width scaling

The issue of what constitutes a tolerable ELM energy loss (WELM) with respect to erosion/damageof plasma-facing components is evolving and is a function, particularly in the divertor, of the choiceof monoblock front surface shaping and of the response of W material to sustained operation at hightemperatures in the presence of multiple, repetitive transients. The latter can be far away frommelting, but still lead to surface cracking which may later seed the development of deeper macrocracks.
The original specification adopted in the ITER Heat and Nuclear Load Specifications[ITER_D_2LULDH, 2009] for the allowed ELM energy loss, WELM, was derived from plasma
gun experiments in which melting of the edges of perfectly aligned, castellated tungsten targets wasobserved for energy densities above ~0.4 MJm-2 in plasma pulses with temporal envelopes
characteristic of those expected during uncontrolled type-I ELMs in ITER [Klimov, 2009].
Together with the conservative assumption of no broadening on the divertor target during the ELM,
and the specification (derived from sparse experimental data) of asymmetric energy sharing
favouring the inner target, this led to a requirement for WELM ≲ 0.6 MJ.
The above criterion was established on the basis of power flow to an axisymmetric, unshaped
divertor target, before the fine details of the shaping required on the divertor monoblocks wereaddressed. Extensive studies of various shaping options have now been performed, accounting for3-D ion Larmor orbits [Gunn, 2017]. During ELMs, when Larmor radii are on the mm scale,toroidal gap loading is predicted to occur which cannot be completely prevented, at both inner andouter targets, by surface shaping. The consequences of repeated flash melting or excursions well
above the recrystallization temperature of these edges are unknown at present, but gradual erosion
cannot be excluded. If this would eventually be considered intolerable for ITER operation, then it
could be that, for example, divertor monoblock toroidal gap edge melting may set a limit on theallowable ELM size. Furthermore, the point at which this limit may have to be respected could
occur earlier in the operational campaigns than previously thought.
It may even be the case that, if the simple theoretical model offered in support of the parallel ELM
energy density scaling [Eich, 2017] (see Appendix C) applies to ITER, mitigating ELMs may notsignificantly reduce the peak ELM target energy densities (simply the area over which the energy isdeposited on the target varies – this is observed experimentally), so that only full suppression would
avoid the ELM-induced melting problem at the divertor. One should also bear in mind thatmonoblock shaping always increases the stationary power flux density onto the surface and the42ITR-18-003

increasingly recognized need to remain below tungsten recrystallization temperature, including
ELM driven temperature excursions (see 2.2.6), may ultimately place tighter controls on theallowed power flux densities in the high performance phases.
ELM interactions with the main wall have not received the same level of scrutiny as the divertor
interactions. On ITER, however, their effects on erosion can be significant under certain conditions,
due notably to the much lower melting temperature of Be compared with W. Depending on thecharacteristics of the ELM filaments, SOL transport and the design of the magnetic equilibrium(e.g. [Kocan, 2015]), which determines the fraction of ELM energy reaching the walls, significant
melting could occur in the secondary X-point region for large WELM. This is not expected to be a
problem until burning plasma operation [Loarte, 2008], though the detailed calculations of gap
loading, which have been performed for the divertor, have not yet been attempted for castellations
on the first wall. Even if some degree of Be tile gap melting does occur, it is expected to be lessproblematic, due both to the higher tolerance of the plasma to Be impurity and to the fact thatstationary handling of power fluxes on the blanket first wall is not as critical as for the divertor.
2.2.4

Ammonia formation during N2 seeding

Ammonia (NH3) can be formed in tokamaks when nitrogen is seeded into the SOL/divertor plasma
to reduce divertor heat fluxes. Experiments in the all-metal JET and ASDEX-Upgrade tokamaks
have shown that a few percent (up to 8% in AUG [Neuwirth, 2012]) of the injected nitrogen isconverted into ammonia. Tritiated ammonia is efficiently retained on the active charcoal used in the
ITER cryopumps; it can only be recovered with full regeneration of the cryopumps. The tritium
inventory on the cryopumps cannot exceed 180 g and regular regenerations (to ~100 K) will be
performed during operations to recover the pumped hydrogenic species, a process which requires~600 s, but which would not allow for ammonia recovery. Regenerations at higher temperature,whilst of course possible, are very time-consuming (about 6 hr per pump, 6 pumps in ITER) andtherefore significant ammonia production could strongly affect the ITER duty cycle. The formationof tritiated ammonia also has an impact on the design of the tritium plant which needs to be able todecompose ammonia to recover tritium and at the same time avoid the formation of deleterious
nitrogen oxides.
While it is clear that ammonia formation proceeds through surface reactions between nitrogen andhydrogenic radicals [ITER_D_QFF9XY, 2015], it is currently unclear whether these reactions
proceed on plasma-exposed or plasma-shadowed areas and it is, therefore, currently not possible topredict the ammonia formation rates to be expected in ITER. Experimental activities are ongoing intokamaks and laboratory devices to elucidate the formation process of ammonia under fusionrelevant conditions and to determine the reaction rates of the elementary steps involved.
Continuation of this effort is required to allow the inclusion of the relevant nitrogen chemistry in
fluid plasma-boundary codes.
2.2.5

Differences in medium-Z impurity seeding (N2 vs Ne)

With the emergence of major tokamak experiments with all-metal walls in recent years (notably
JET and ASDEX-Upgrade), it has become increasingly obvious that the plasma performance foundin highly radiative scenarios with all-carbon PFCs is not always easy to reproduce when extrinsic
seeding is used to replace the natural carbon impurity for divertor power flux control.
Nitrogen seeding provides the best performance with respect to older results in an all-carbon
environment – not surprising considering the proximity of the radiation functions between C and N.
As discussed in 2.2.4, however, the formation of tritiated ammonia during nitrogen injection is43

ITR-18-003

likely to be an issue for ITER, mostly in the sense of a limitation on the achievable duty cycle.
Neon would thus be a preferable alternative, with the added benefit that being a fully recycling gas
naturally extends the timescales for avoidance of divertor monoblock overheating in the event ofloss of seeding or other divertor reattachment events. This said, recent results from dedicatedexperiments on JET indicate that saturation of wall surfaces with nitrogen can result in rather longtimescales for loss of detachment in the event that seeding gas is removed.
There is a fundamental difference between ITER and machines such as ASDEX-Upgrade and JET
in the sense that machine size, magnetic field and achievable pedestal temperatures are insufficient
to provide the combination of high divertor compression and low main chamber radiation for Ne inpresent devices. Experiments find in general [Bernert, 2017; Giroud, 2014] that radiation in N2seeded plasmas occurs preferentially in the divertor, whilst with Ne, for the same fuelling rates,
main chamber radiation increases and there is little or no increase in divertor radiation. The
increased main chamber radiation lowers the power crossing the separatrix and affects ELM
frequencies. In addition, pedestal transport is modified by the presence of impurities. Nitrogen tends
to increase pedestal top temperatures and hence confinement in both devices, whilst Ne injection
increases the pedestal top density in ASDEX-Upgrade and reduces the pedestal pressure
significantly on JET. In both devices, Ne injection beyond a certain level leads to an unstablesituation.
In ASDEX-Upgrade, the decrease in ELM frequency reduces the effect of ELM impurity flushingand the increased pedestal density gradient enhances both Ne and W inward transport in-between
ELMs, further enhancing core radiation losses [Bernert, 2017]. Both effects, together with the factthat Ne does not allow a reduction in divertor power loads by increased SOL/divertor radiation,mean that power exhaust with Ne appears difficult to study in ASDEX-Upgrade, even though thedevice is operating at values of Psep/R close to those expected in burning plasmas on ITER. In JET,
with values of Psep/R currently about a factor of 2 lower than on ASDEX-Upgrade, Ne tends toreduce the power flow across the separatrix, provoking H-L back transitions, making a propercomparison between N and Ne difficult until more heating power is available [Bernert, 2017;
Giroud, 2014].
These results may be contrasted with those found on radiative divertor experiments on Alcator CMod with molybdenum PFCs at high magnetic field [Loarte, 2011], in which for both N2 and Ne
seeding in EDA modes, partially detached divertor conditions were achieved at normalized
confinement, H98 ~1 with edge power flows only marginally above the H-mode threshold power (byfactors of 1 – 1.4), as in ITER. Here, as in the JET and ASDEX-Upgrade experiments, at highradiative fractions, the radiation is observed to occur mostly in the X-point region, with someincursion onto flux surfaces just inside the separatrix. Apparently, in C-Mod, this was insufficient tostrongly affect the pedestal top, even in the case of Ne.
It is extremely difficult at present to confidently extrapolate the findings on current devices to ITER
with respect to performance with low-Z impurity seeding. Simulations with SOLPS for burning
plasma conditions [Pitts, 2015] show that for equivalent SOL power flow, ITER will compress both
Ne and N strongly in the divertor, in the strike point regions, albeit with the Ne radiation distributed
over a slightly more extended volume. Similar partially detached conditions are achieved, thoughdifferent seeding rates are of course required in both cases. Integrated modelling using the boundary
plasma scaling approach (see [Pacher, 2015] and Appendix C) has been applied to show that bothN2 and Ne provide adequate burning plasma operational windows, with N2 marginally better due toslightly lower core radiation in comparison with Ne. However, more work is required here to44ITR-18-003

improve the integrated model and to extend the N2 seeded boundary cases which are so far limitedin scope (Ne seeding has traditionally been the favoured approach at ITER for power flux control).
In the end, it is likely that the large physical size of ITER and the much higher expectedpedestal/separatrix temperatures in comparison with current machines mean that the differences
being found today between N2 and Ne will disappear at the ITER scale. Higher pedestal/separatrix
temperature means lower Ne radiation in the core plasma. In the SOL/divertor, mean-free paths forionization of neutral Ne are larger than for N and the relative importance of thermal and friction
forces differ for the two species, with the net parallel force in the divertor being lower for Ne thanN. Both effects drive lower divertor impurity enrichment for Ne, but the differences are likely to bemuch reduced with increasing device size.
Both experimentally and theoretically, the issue of low-Z impurity seeding is an area of very active
research and significant advances can be expected in the coming years. It will be of particularimportance to demonstrate, if possible, scenarios in which ELM control is properly integrated withextrinsic seeding for power flux control at high performance. Further experiments on JET, at thehighest possible input power and current for which pedestal temperatures will be closest to those
expected in ITER, have the best chance of reducing the gap to ITER in terms of operating
conditions (size, performance) by the comparison of the efficiency of N2 and Ne seeding withregard to divertor power flux control compatible with acceptable core plasma performance.
2.2.6

Tungsten-related material issues

One of the frequently mentioned advantages of tungsten as a plasma-facing material is its highmelting point. Much lower temperatures are, however, required for recrystallization/grain growth tooccur, resulting in strong changes in the material mechanical properties and, in particular, a
reduction in strength/hardness, and shock resistance despite an increase in ductility. Appearance ofmacro-cracks on W during high heat flux testing at 20 MWm-2 is indeed attributed to exhaustion ofductility of the recrystallized W - the temperature being high enough for recrystallization to occur.
To avoid the deleterious effects associated with recrystallization, an operational budget will need to
be defined accounting for the time/temperature dependence of the recrystallization process (seeAppendix C). Little is known about the recrystallization kinetics of the actual W to be used for the
ITER divertor, and more information will be required to consolidate the actual operational budget.
Even if bulk recrystallization is avoided, significant surface damage can be caused by ELMs.
Surface roughening/cracking/melting can occur depending on the ELM energy density. The
evolution of the surface damage and its dependence on the base material temperature, the energy
density and the pulse number has been extensively studied in electron-beam facilities and using
powerful pulsed plasma and/or laser systems (see for example [Wirtz, 2017]). Most of theseinvestigations have been performed on pristine material and in the absence of any particle
irradiation.
Some studies have revealed the existence of synergetic effects for combined particle/heat loadingwhere the general tendency shows a lower damage threshold for a plasma-exposed surface. A big
unknown for ITER is whether and how these effects might affect the material compatibility withhigh performance plasmas. In particular, it is at present impossible to define a tolerable level ofdamage before deteriorated material properties would start to impact plasma performance due toenhanced tungsten release or reduced power handling capabilities. This is clearly an area where
further research is needed before the start of ITER operations.
The effect of long-term stationary and transient plasma exposure (be it H, D or He) on the thermomechanical properties of W clearly remains an important topic where little information is available,
45

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although indications exist for the modifications of material properties following plasma exposure.
Extrapolation of these observations to the very high fluence expected in ITER is currently too
uncertain. A proper assessment of the impact for ITER requires that these effects are accounted forin the modelling of the thermo-mechanical response of the material (as performed in [Panayotis,
2017] for example) and in turn would require knowledge of the evolution of material properties
with plasma fluence (ultimate strength, toughness, thermal conductivity, etc.).
2.2.7

Castellation gap heat loads

Since the last major version of the IRP, a great deal of progress has been made on the divertor
design, including the elimination, in 2011, of the first carbon/tungsten divertor from the ITER
operational plans [Pitts, 2013]. This decision required an acceleration of the design work for the
full-W version, which must now be in place for PFPO-1 and must guarantee operation well into the
FPO phases (see 4.4 for more on the lifetime requirements and possible divertor upgrades).
One important design area that had not been addressed for the full-W divertor was the question of
W monoblock shaping. Due to the high stationary power flux densities of which ITER will becapable, and the potential for significant transient energy densities, melting at leading edges acrossgaps between neighbouring monoblocks that arise from manufacturing and assembly tolerances
becomes a real issue for component lifetime in the case of metal divertor targets. As a result of
several years of design and physics studies (see e.g. [Pitts, 2017; Gunn, 2017; Gunn, 2016; Hirai,
2018]), the final PFC shaping is now essentially decided, for both the high heat flux and baffle
regions of the vertical targets. What these physics studies have highlighted is the very complex
nature of plasma interactions in the vicinity of gaps between PFCs; when the magnetic field anddivertor geometries and ion Larmor orbits are properly accounted for, gap edges (running in both
poloidal and toroidal directions) can be accessed by plasma ions when otherwise they would
normally be magnetically shadowed under a pure optical approximation. This is particularly true
during ELM transients, when energetic ions arrive at the divertor targets.
Sophisticated monoblock surface shaping; for example, the very complex, high precision machining
applied to the individual components in the JET bulk W outer divertor target [Mertens, 2011] is nota realistic option for the ITER divertor design. The reference ITER design is a simple toroidal bevel
to protect misalignment of toroidally adjacent monoblocks in the high heat flux areas [Pitts, 2017].
This will not protect loading of gaps running in the toroidal direction between poloidally
neighbouring blocks. However, as discussed in Appendix C, any applied shaping reduces theallowed parallel power flux density which can be intercepted by the component; for the particularcase of ITER, the current baseline toroidal bevel reduces the allowable stationary heat flux density
by ~40% compared to a non-shaped solution and places strong limits on the allowed stored energy
loss per ELM for the avoidance of toroidal gap edge melting. Toroidal gaps can be protected withthe addition of a further poloidal bevel, but this only works at the outer target in ITER and will add
a further penalty on the allowed stationary parallel heat flux densities.
The many thousands of gap edges in the ITER divertor mean that careful attention will have to be
paid to power loading as performance improves during operations. Gap edge loading occurs on the
Larmor radius scale, typically on the order of a few hundred microns, so that only the highresolution divertor IR systems viewing a small region of the vertical targets at single toroidal
locations will be appropriate for the detection of these interactions. The lower resolution, machine
protection IR systems will, however, be capable of detecting the stationary top surface loading on asubstantial fraction of the individual monoblocks in the strike point areas. Although the ITER
divertor design requirements have stimulated a good deal of recent R&D on current devices to study
leading edge and gap interactions, this work should continue, in particular aiming to directly verify
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the existence of ELM driven ion orbit loading. In this context, it is also important to continue
theoretical studies of ELM evolution and transport to provide greater confidence in thespecifications for particle energies and time envelopes used to derive the required ITER monoblock
shaping.
2.2.8Radio-Frequency-assisted wall conditioning techniques

The fact that the toroidal magnetic field, generated by superconducting magnets, will be
continuously maintained during ITER operational periods, drastically reduces the efficiency of DC
glow discharges. Ion Cyclotron Wall Conditioning (ICWC), operable in the presence of the toroidal
field, is therefore foreseen for fuel and impurity removal between plasma pulses. This will also
contribute to the mitigation of the tritium inventory build-up in the machine in the FPO campaigns.
In case of unavailability of the ICRF system in PFPO-1, Electron Cyclotron Wall Conditioning(ECWC), also compatible with Bt, could be employed. Both methods use the conventional heating
antennas to produce low temperature magnetized discharges. In ICWC, coupling of the power to thedischarge results principally from collisional absorption of the radio-frequency (RF) energy by theelectrons (non-resonant coupling), and the discharges can therefore be produced at any Bt. ICWC
discharges present poloidal and radial non-uniformities inherent to RF wave absorption schemes.
Operation in monopole phasing and/or mode conversion schemes [Lyssoivan, 2012] have improved
discharge homogeneity and optimized wall coverage. ICWC is expected to require 1–5 MW ofpower in ITER with a duty cycle 1 s on/30 s off to make a He, H or D plasma with ne ~ 1017–
1018 m−3 and Te ~ 5–10 eV [Douai, 2011].
ECWC discharges can only be produced in the presence of the Electron Cyclotron Resonance layer(ECR) inside the vacuum vessel. As in ICWC, but to a greater extent, localized power absorption
and high density in ECWC plasmas (ne ∼ 1018-1019 m-3) can affect discharge uniformity andcleaning efficiency, though pulsed operation is used to mitigate the latter effect. However, thetechnique suffers from low absorption per pass in the optically thin ECWC plasma, even at thefundamental of the ECR frequency, leading to the risk of exposing PFCs or diagnostics to excessive
stray radiation. The ECWC operational parameters (pressure, power) should, however, be very
similar to those of the ECRH-assisted breakdown foreseen in ITER. ECWC may require 1–10 MW
at 170 GHz, with a duty cycle similar to that of ICWC.
In both techniques, using H or D as working gas the main flux to the wall is that of isotropic neutralhydrogenic species, either desorbed from wall surfaces [Douai, 2011] or resulting from moleculardissociation. The ion flux parallel to the magnetic field lines is largest on limiting surfaces, reachingtypically 1021 ions m-2s-1, and decreases exponentially in the plasma edge until a few 1019 ions m-2s1[Lyssoivan, 2012]. For comparison, ion fluxes in a GDC plasma are typically 2-4 orders ofmagnitude lower (~1017 ions·m-2s-1) [Kogut, 2015]. In ICWC, fast charge exchange neutrals can in
addition be produced, with temperatures above 1 keV and energies up to 50 keV. The large
penetration depths of these particles into sub-surfaces (up to 200 nm in Be) makes them particularly
attractive for T-removal.
ICWC is routinely applied between shots in the KSTAR tokamak for impurity removal [Kwon,
2011], but it is mainly considered for ITER as an efficient option to mitigate the T inventory buildup. Present extrapolations from current devices to ITER, in particular results from experiments in
JET with ITER-like wall (JET-ILW) (operated with walls at 200°C), indicate that ICWC could
remove up to ~0.35–0.6 g of T between plasma shots, i.e. an amount comparable to the estimatedretention per pulse [Brezinsek, 2013-1; Shimada, 2011]. Experimental and modelling data [Kogut,
2016], however, show that surface temperature has a non-negligible impact on the isotopic

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exchange efficiency. New experiments in the JET at lower wall temperatures (ICWC will be
operated at 70°C in ITER) would therefore appear to be particularly relevant in this respect.
The efficiency of ECWC with metallic PFCs is less well documented. Recovery from disruptions
with ECWC only has been successfully demonstrated on JT60-U [Itami, 2009]. Even in devices
with carbon PFCs, the scarce published data on fuel removal in the literature indicates a lower
efficiency than ICWC [Douai, 2016].
2.2.9

Divertor detachment control

The control of divertor power loading will be one of the most critical aspects of ITER operation andwill need to be robust and reliable. Although the W divertor monoblock technology, which ITER
will use on the vertical targets, is the most advanced currently available and can operate with steadystate surface loads of ~10 MWm-2, even relatively short durations above this value must be limited
if recrystallization is to be avoided (see Appendix C and section 2.2.6). Stronger excursions,
resulting in global surface melting in quasi steady state (namely on timescales significantly longer
than ELM or disruption transients), will have severe consequences for surface topologicaldeformation and must be absolutely prevented. The ultimate limit (defined by the achievement of
the critical heat flux at the cooling interface [Escourbiac, 2016]), if sustained for long enough (ofthe order of a few seconds), can lead to a water leak requiring immediate and lengthy in-vesselintervention and the replacement of an entire divertor cassette. Although the heat flux densities
required for this to occur can only be achieved during burning plasma operation at high current
when SOL input powers are sufficiently high, Appendix C demonstrates that heat flux control may
already be necessary during operation in PFPO-2, when PSOL may attain values >40 MW. It will inany case be mandatory to develop and routinely deploy detachment control in the pre-nuclear
operational phases before the introduction of D and T and the activation of components.
It is only relatively recently that power flux control has begun to be seriously developed in current
devices, where the use of W divertor targets, invariably inertially cooled (e.g. ASDEX-Upgrade andJET), means that without it high power operation is not possible, even for relatively short pulse
durations on the order of seconds. With several medium-sized superconducting divertor devices
now in operation, of which two (EAST and WEST) use actively cooled W divertors and a third(KSTAR) has plans to upgrade to full-W actively cooled components by 2021, progress incontrolling heat loads can be expected to become much more widespread.
To date, the most robust detachment controller has been that developed on ASDEX-Upgrade, where
the current from an outer target shunt (with appropriate real time processing to remove ELM spikes)
is used to measure the thermoelectric current into the outer divertor, which, because the inner target
Te is always very low between ELMs, is a good approximation to the outer target T e and hence ameasure of detachment state [Kallenbach, 2010]. The signal is used inside a PI controller todetermine the impurity gas puffing rate required to achieve a given set divertor Te. The system
works routinely with nitrogen, which is well compressed in the ASDEX-Upgrade divertor (see2.2.5) and is thus a direct actuator for divertor radiation. More recently, a real time double radiative
feedback technique has been developed on ASDEX-Upgrade to simultaneously control injection ofa main chamber radiator (Ar) and N in the divertor [Kallenbach, 2012]. Here a subset of 3 foilbolometer channels are used in the main chamber together with a fourth in the outer divertor leg, orin conjunction with the thermoelectric current controller. Other tokamaks have previously tried
feedback on bolometer chords (e.g. [Goetz, 1999; Asakura, 2009]). A real time nitrogen feedback
control technique has also been tested in JET using the degree of detachment obtained from ELMfiltered outer target Langmuir probe ion current densities [Guillemaut, 2017]. An earlier seeding
control attempt on JET used a VUV nitrogen line [Maddison, 2011].
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Unfortunately, owing to the complex structure of the water-cooled monoblock plasma-facing units,
divertor shunts cannot be incorporated into the ITER divertor targets and thus power flux control
methods must be developed based on the sensors which will be available. As well as an extensive
array of bolometer channels in both main chamber and divertor, ITER will be equipped withcomprehensive IR thermography, covering ~90% of the high heat flux areas of the divertor target.
The spatial resolution will be rather coarse (for example on the scale of a single monoblock) inmany regions, but since divertor target loading is rather toroidally symmetric in comparison withthe first wall, the zones with highest spatial resolution can be preferentially chosen. However, IR
surface power measurements have not to date been used for heat flux control in tokamaks, whichmeans that the community has not yet developed a feedback scheme on the key parameter ofinterest for ITER: the surface heat flux density. There are issues associated with measuring power
fluxes in detaching/recombining divertor plasmas (e.g. intense bremsstrahlung radiation in the IRband, strong reflections in metallic environments) and with the reliability of IR cameras operatingin the 3-5 m band. Research has typically focused on more robust, simpler and usually indirecttechniques, such as the feedback on bolometer channels mentioned above. Nevertheless, theextensive IR coverage of the divertor being installed for ITER means that efforts should be
intensified in the R&D Community to demonstrate that real time, IR-based target heat flux control
is possible for metallic components in a highly dissipative (and hence radiative) regime, and toproperly assess the limitations and/or find appropriate workarounds. Some effort on thedevelopment of a real time capable IR system has begun at ASDEX-Upgrade [Sieglin, 2015] and is
expected to be tested in the near future.
Additional sensors with which ITER will be equipped and which may form part of detachment
control loops are the arrays of divertor Langmuir probes and thermocouples attached to monoblocks
in several toroidal locations. Since divertor neutral pressure is a key parameter in determining thedegree of detachment, it is also possible that the ASDEX-type fast neutral pressure gauges that will
be installed at various locations in several ITER divertor cassettes could eventually be used in acontrol loop. Nevertheless, issues of lifetime are important here: in-vessel diagnostics attached todivertor targets and requiring extensive electrical cabling to connect them with elements outside thevessel may not survive long enough to be considered as robust for heat flux control in the nuclear
phases, when it will be needed routinely and during which operating conditions in the divertor willbe the most severe.
2.2.10 PWI/plasma boundary code validation
In addition to the experimental efforts mentioned above, it is important that by the time ITER
operation starts, the modelling tools in use within the community for plasma-wall interaction andplasma boundary physics studies in ITER-relevant regimes have reached a sufficient level ofmaturity and validation against experiments, including cross-benchmarking demonstrations[Wiesen, 2017]. Although it could be said that 2-D edge plasma fluid codes have, for the most part,
already reached this level of maturity (e.g. [Kukushkin, 2011]), some questions remain about thedifficulties that many such codes encounter when attempting to reproduce features such as electricfields [Chankin, 2007] and plasma flows in the SOL [Asakura, 2007], or the inclusion of drifts at
the ITER scale. It is believed that many of these shortcomings are a result of the implicit
Maxwellian fluid assumption used for the plasma description in these codes and could only be
addressed by explicitly including plasma kinetic effects [Stacey, 2016]. Purely kinetic approaches[Takizuka, 2017], however, require enormous computer resources to provide results at the ITER
scale.

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Although it is expected that computing speeds will continue to increase in the coming years, it isoften the case that any improvement in computer performance is balanced by further refinements ofthe physics model in the codes, keeping the run-times roughly constant. However, during ITER
operation, there will be a strong need for a fast and reliable plasma edge solver, coupled to asuitable core model, which can be used for online analysis and preparation of ITER discharges.
Fully consistent core-edge coupling models, however, are only now beginning to be established[Falchetto, 2014]. Moreover, the use, for example, of magnetic perturbations for ELM mitigation
ultimately requires a 3-D description of the ITER boundary plasma. Current tools [Feng, 2017] still
struggle to reach the highly dissipative regime present in the partially detached ITER divertor
plasma, and significant effort will need to be devoted to overcome this hurdle. Conventional (i.e.
non-turbulent) transport codes will also need to find a way to properly incorporate a turbulenceansatz [Rognlien, 2004] to obtain, for example, a more realistic description of the blobby
intermittent transport which is experimentally known to dominate far SOL transport and, therefore,
governs the particle and heat loads received by the main chamber walls. With a correct descriptionof the fluxes reaching the walls, either through 2-D or 3-D models covering the entire vacuum
vessel volume, more precise modelling of wall material migration, redeposition patterns and fuelretention in the plasma-facing components will become possible [Baelmans, 2011].
A great deal of development has taken place in the area of material migration modelling tounderstand erosion/deposition effects and to improve the accuracy of fuel retention estimates. Twotypes of approaches have been followed to study either local effects [Borodin, 2011] at the level ofa given component or a given region of the machine, or global poloidally-resolved migration[Schmid, 2015]. The former approach allows the 3-D geometry of the ITER first wall panels (FWP)
and, therefore the effect of component shaping on local material redistribution, to be accounted for.
The latter approach has been shown to successfully reproduce the retention rates measured by gas
balance in JET and the campaign averaged erosion/deposition pattern. The same modelling, whenapplied to ITER, shows that co-deposition of T with Be could occur predominantly either in thedivertor or on the first wall itself, depending on the nature of the far-SOL plasma profiles. Three
dimensional effects could play a strong role here given the highly shaped and conformal ITER firstwall. The development of global 3-D material migration models would therefore allow a better
description of erosion/deposition processes in ITER. This, in turn, clearly links to the developmentof 3-D boundary SOL codes mentioned above which can provide the plasma background used to
study impurity generation and transport.
Migration modelling aims at describing where eroded particles deposit in the vessel and how muchtritium is trapped in the resulting deposited layers, particularly in the case of ITER which will use a
low-Z first wall with low erosion energy threshold for ion impact. One-dimensional
diffusion/trapping models have been used [De Temmerman, 2017] to study the efficiency of tritium
removal from co-deposits during PFC bake-out. The 1-D approach does not permit a consistentmodel describing the full outgassing process in which different regions of the machine are heated
up to different temperatures (240o C for the first wall and 350o C for the divertor are considered forITER) or the fact that the tritium retention in co-deposits will be strongly spatially dependent on thelocal co-deposition conditions [De Temmerman, 2008]. A possibility to tackle this issue is throughthe coupling of a diffusion/trapping model to the above-mentioned migration codes.
The area of plasma-material interactions represent a particular challenge for modelling because of
the large span of temporal and spatial scales involved in the full process. Going from the atomistic
scale of the ion/surface interaction processes (and resulting defect creation, trapping, diffusion, etc.)
to the impact of these processes on the mechanical properties of the plasma-facing material, (i.e.
being able to predict how the material thermal and mechanical properties will evolve with time50

ITR-18-003

during operations), requires the development of multi-scale models which currently do not exist andare far out of reach. Some efforts are ongoing to bridge different scales and describe the interactionsof hydrogen and helium isotopes with tungsten surfaces and the initial stages of morphology
changes that they induce [Lu, 2014; Wirth, 2015], but a substantial expansion of these efforts will
be required. The significant advances taking place in parallel on the experimental side such as theuse of micro-mechanical measurements and the development of in-situ diagnostic systems will help
by providing data to validate multi-scale models as they are being developed. All of these activities
will need to be coupled together [Brooks, 2014] and embedded in the ITER integrated modelling
framework, IMAS (Integrated Modelling and Analysis Suite [Imbeaux, 2015; Pinches, 2016]) inorder to build reliable workflows able to provide a consistent picture of the ITER plasma andsurrounding plasma facing components conditions.

2.3 Plasma Disruption Management
Plasma disruptions can cause substantial heat loads to the divertor and the first wall, high electromagnetic loads on in-vessel components and on the vacuum vessel itself, and they may also lead tothe formation of runaway electrons (REs) that can potentially cause water leaks due to the very
localized deposition of energy on first wall panels or plasma-facing components (PFCs) of thedivertor. Operation of ITER will have to strongly focus on avoiding disruptions with a high successrate and on mitigating those in which avoidance techniques fail [de Vries, 2016-02; Lehnen, 2016;
Hollmann, 2015; Lehnen, 2015-1]. The Research Plan takes this requirement into account andsignificant operational time for commissioning and optimizing the disruption mitigation system andfor establishing appropriate disruption forecasting and avoidance techniques and schemes, linked
closely to basic and advanced control systems, is reserved in all three operational phases, PFPO-1/2 and FPO. This research is explained in greater detail in the following sections. When possible, theResearch Plan aims to commission each system one operational period before it will be needed.
Figure 2.3-1 shows the various limits for melting of PFCs and for electro-magnetic loads during
disruptions in ITER. It also indicates the potential maximum energy at various plasma currents thatcan be deposited by REs in case of full magnetic energy conversion, which is the most pessimistic
scenario. The assessment of thresholds for PFC melting or for a possible failure of cooling channels
after runaway impact is an ongoing R&D activity. Present predictions of these thresholds have large
ranges due to uncertainties in the affected surface area. However, the foreseen energies at a plasma
current Ip = 7.5 MA are presently expected to be sufficient to cause water leaks. It is important tonote that within the green area in Figure 2.3-1, in which no PFC melting is predicted to occur,
runaway electrons (RE) formation is not expected for unmitigated disruptions in the non-active
phase of operation. In these conditions no influx of radiating impurities is expected to take place(moderate beryllium influxes are not expected to cool down the plasma significantly) so that highelectric fields during the current quench will not develop and, in addition, no seed REs from tritiumor Compton scattering are present in these plasmas. When mitigating disruptions, it is the role of the
disruption mitigation system (DMS) to ensure RE avoidance by an appropriate choice of theinjected species and quantity.
Given the high impact of disruptions on component lifetime and machine availability, anappropriate management scheme has to be in place to allow the timely achievement of theobjectives of the ITER Research Plan. The disruption loads and their consequences determine theallowable disruption rates (and consequently the necessary avoidance rates) and mitigation successrates for the various phases of operation with increasing thermal and magnetic energies towards the
Q=10 baseline scenario. To define these rates, a disruption budget consumption (DBC) that
represents the ‘costs’ of a disruption at a specific magnetic and thermal energy is introduced
51

ITR-18-003[Lehnen, 2016]. This DBC can be attributed to any disruption impact that is cumulative or forwhich an occurrence probability can be defined. Specifying the DBC values with a good physics
basis is a prerequisite for establishing a reliable strategy for disruption management. Improving thephysics basis for the definition of the DBC is a continuous process and, especially during operation,
the forecast and the accounting of the DBC will have to be updated based both on the postdisruption assessment of the actual impact that occurred and on scalings for disruption loads thatwill be established. This careful accounting will allow identifying deviations from the planned
budget consumption and – if required – to react accordingly by adjusting the operational plan. It is
important to also note that mitigated disruptions can cause accumulation of the DBC for example
through radiation flash heating of first wall components or through consumption of fatigue lifetimeof in-vessel components. A main risk for the execution of the research plan is the accidental
generation of runaway electrons during disruption mitigation. Since extrapolation from a certain
plasma current level to the next has high uncertainties, runaway avoidance during disruption
mitigation has to be carefully validated throughout the approach towards 15 MA operation.

Figure 2.3-1: Operational range of ITER in terms of thermal energy and plasma current. Thermal load
limits during the thermal quench are indicated by horizontal dotted lines and during the current quench by
the vertical dotted line. At high plasma currents, it is at present assumed that cat. III halo currents will
occur if electro magnetic (EM) load mitigation is not activated. The green area indicates the parameter
range for which melt damage from unmitigated disruptions is not expected. The symbols indicate thetarget values for the planned operational scenarios. Operation at high current without auxiliary heating is
not planned, but energies would already exceed melt limits under such conditions (orange line). The RE
energy shown along the top axis is the maximum possible deposited energy for a given plasma current
assuming full conversion of the magnetic energy to kinetic energy.

The maximum downtime after activation of the DMS is defined to be 3 hours. Although the highest
priority is on mitigating the direct consequences of a disruption, the optimization of the DMS has
also to take into account the impact of the mitigation action on the pulse rate and possibly also on
52

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the performance of the subsequent pulse. Therefore disruption rates – and also mitigation rates inthe early phase of operation – will have to be limited, but also the injected species and quantities
have to be optimized to be not unnecessarily excessive, which would cause delays due to longer
vessel pump-out times. Early verification of the impact of high material injection quantities on
vacuum systems like cryopumps and gas handling systems is mandatory. This can be mainly done
without plasma operation. Conditioning after DMS activation may be needed to remove residual
impurities in the first wall, neutral beam injection (NBI) beam duct, etc. At present no quantitative
analysis of this effect is available. In case of unmitigated disruptions, a robust conditioning pulse
might be required in case of degradation of the first wall or divertor conditions to support highperformance plasma operation through for example material splashing. These losses of operationaltime have to be taken into account and will also contribute to defining targets for disruption rates.
This will also require reducing the number of mitigated disruptions in plasma parameter ranges forwhich no damage on PFCs is expected. Moreover, it will have to be considered when setting targets
for the false alarm rates of the disruption predictor.

2.4 First Plasma
The operational activities surrounding the production of First Plasma in ITER, targeted within theStaged Approach for December 2025, consist of 3 main phases: Integrated Commissioning, First
Plasma and Engineering Operation.
2.4.1

Integrated Commissioning

As installation of the various ITER plant and auxiliary systems is completed, each of the systems
will undergo instrumentation and control (I&C) integration with ITER CODAC and systemcommissioning. This should ensure that each system satisfies those design requirements that can betested in the absence of plasma in advance of their integration into the tokamak operations
framework. Following completion of the construction of the tokamak core, scheduled for late 2024,
a first period of Integrated Commissioning, lasting ~12 months, will occur, with the achievement ofhigh vacuum conditions in the vacuum vessel, cooling of the superconducting magnets anddemonstration of a significant level of performance, together with the commissioning of all major
tokamak subsystems required to support plasma operation. A schematic of the sequence of eventsand estimates of time required for the various commissioning activities in shown in Figure 2.4-1.
Although a preliminary flow and planning for Integrated Commissioning have been prepared within
the framework of earlier schedule exercises, a new activity has been launched to develop an up-todate analysis of this phase of operations in order to develop improved estimates of the time requiredfor completion of the various activities within this period and to confirm the detailed sequence ofactivities. The major activities determining the critical path through Integrated Commissioning are
illustrated in the figure: the vacuum vessel and cryostat enclosures are evacuated and then leakdetection is made at room temperature; after circulating, filtering and heating the water, the vacuum
vessel and all in-vessel components are baked at high temperature and a second leak detection is
performed; cooling of the superconducting coils to cryogenic temperatures and then energizing andcharging of the coils are performed to approximately 50% of their maximum current – this involves
extensive operation of the coils individually and then as an integrated system, together withcommissioning of the protection systems. Finally glow discharge cleaning of the vacuum
components is carried out to the extent necessary to achieve reliable plasma operation.
Integrated commissioning of control and auxiliary systems required for First Plasma is also
performed during this period. The Gas Injection System (GIS) is likely to be the first auxiliary53

ITR-18-003

system to undergo integrated commissioning, since it will be used during glow discharge cleaningof the vacuum vessel. Following offline commissioning of the ECRH power sources andtransmission lines, the ECRH system will be integrated into the operational systems and short-pulse
injection (up to 100 ms) will be used to condition one upper launcher. A subset of Diagnostics havebeen identified as essential for First Plasma (see Table H-1 ) and these must be fully commissioned
to be ready to take data during the First Plasma campaign – the magnetics diagnostic will be criticalto the implementation of this campaign and calibration and measurement consistency tests during
this period will be an essential activity. The Plasma Control System, Central Interlock System andCentral Safety System will also undergo integrated testing under CODAC during this period (PCS
control of the poloidal field (PF) and central solenoid (CS) currents will be implemented during
Magnet commissioning).
The culmination of the Integrated Commissioning activity would be a ‘dress rehearsal’ for plasma
operation of the tokamak, in which all tokamak, plant and auxiliary systems are operated under
CODAC control and satisfy the respective requirements for plasma operation.

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ITR-18-003

Figure 2.4-1: Schematic of thesequence of events during:
Integrated Commissioning
First PlasmaEngineering Operation

55

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2.4.2

First Plasma

First Plasma remains, in essence, a demonstration of the successful integration of the tokamak coreand principal plant systems (magnets, power supplies, cooling, cryogenics, vacuum, etc.) and is theconclusion of the first phase of integrated commissioning of the ITER facility. The achievement ofFirst Plasma constitutes a demonstration that all core tokamak systems (i.e. excluding in-vessel
components) have been commissioned successfully and that the device is capable of achieving
plasma operation. During the First Plasma campaign, none of the main in-vessel components will
be installed, in particular no blanket/first wall modules or divertor cassettes. Therefore, the vacuum
vessel, in-vessel wiring, diagnostics and other installed in-vessel components will be protected by
the First Plasma Protection Components (FPPC), consisting principally of 4 poloidal limiters, most
likely with a stainless-steel PFC, capable of supporting 1 MA plasmas at 2.65 T (see, e.g.,[ITER_D_T7KUNB, 2016]). The FPPC also include a divertor replacement structure that protects
the bottom of the vessel, preventing plasma from reaching the divertor region.
Most of the Magnetic diagnostics will be installed for First Plasma, together with a subset of
diagnostics required for the demonstration of plasma operation. This would include, in particular,H monitors, several IR camera lines of site, impurity monitors and hard X-ray monitor for
detection of runaway electrons. Table 2-2 lists the minimum measurement requirements establishedfor First Plasma and Engineering Operation together with the diagnostic systems that will be
installed to meet these requirements (for further discussion, see section H.2 and Table H-1 ).
Several MW of ECRH power will be available to assist the plasma breakdown. This will belaunched from an upper port towards a mirror mounted on the inner wall of the vacuum vessel,
which will reflect the EC radiation back through the centre of the poloidal field null into an opposite
port where a beam dump will be mounted to limit the stray radiation in the torus. A total ECRH
power of 6.7 MW at 170 GHz will be available for breakdown/ burn-through assist.
The essential parameters for First Plasma will be:
Ip ~ 100 kA
tpulse ~ 100 ms
Paux < 6.7 MW (for breakdown/burn-through assist)
Nevertheless, simulations indicate that if burn-through is achieved, a plasma current of > 0.5 MA
might be reached lasting few seconds (see, e.g. [Kavin, 2016-2]).
2.4.3

Engineering Operation

Following First Plasma, a period of 6 months of Engineering Operation will be undertaken with theprincipal aim of commissioning the Magnet systems to full current operation. During this period, if
time permits, the possibility of producing quasi-circular limiter plasmas of up to 1 MA for several
seconds could be explored, since this would provide an opportunity to make initial observations oftokamak plasma operation in ITER prior to Assembly Phase II, and could also provide valuable
information on the operation of the plasma control system in preparation for PFPO-1 operation.
A further key activity which will be implemented during this phase is a careful determination of the(toroidal) magnetic axis of the tokamak. This procedure, which is still under development, is aimed
at obtaining a very precise mapping of the toroidal magnetic field within the vacuum vessel, fromwhich the final position of the shielding blanket modules (to be installed on the vacuum vessel wall)
can be determined to ensure that the individual first wall panels are well aligned
toroidally/poloidally and, most important, radially with the last closed magnetic surface in ITER

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limiter plasma configurations, to optimize their power handling capability during the limiter phaseof the discharge.
Table 2-2 – Overview of minimum measurement requirements and installed diagnostic
systems for the First Plasma and Engineering Operation campaign*
Measurement Requirement

Installed Diagnostic Systems

Magnetics for position, velocity, shape and
MHD mode structure

•
•
•
•
•
•
•

Line averaged electron density (toroidal
polarimeter/ interferometer)• Density Interferometer (single channel)

Runaway electron detection (hard X-rays)• Hard X-ray Monitor

Impurity identification and influxes (visible
and near UV spectroscopy including H andvisible bremsstrahlung), partial systems• H/Visible in EPP12• Vacuum Ultra-Violet Survey• Visible Spectroscopy Reference System(partial) - temporary• X-Ray Crystal Spectrometer

Visible/IR TV viewing (spectroscopicallyfiltered), partial coverage• Visible/IR Equatorial in EP16 (temporary)• Visible/IR Equatorial in EPP12 (partial)

Torus pressure and gas composition (torus
pressure gauges, RGA)

• Pressure Gauges (temporary)

Toroidal Field (TF) Mapping• Temporary set of magnetic pick-ups for TF
mapping

Machine protection• Tokamak Structural Monitoring System• Stray ECRH detector

Magnetics System Electronics & Software
Continuous External Rogowski
Outer Vessel Coils
Steady State Sensors
Flux Loops
Inner Vessel Coils
Diamagnetic Sensors* See Appendix H for further discussion of scope of systems labelled as ‘temporary’ or ‘partial’.

2.5 Pre-Fusion Power Operation Phase (PFPO)The Pre-Fusion Operation Phase will be broken down into PFPO-1, which will exploit up to
20 MW of ECRH power plus an option for 10 MW of ICRF power, and PFPO-2, which willdevelop all three of the auxiliary H&CD systems to reach 20 MW of ECRH power, 20 MW of
ICRF power, and 33 MW of NBI power, in preparation for the Fusion Power Operation (FPO)
phase.
2.5.1

Objectives for the Pre-Fusion Power Operation Phase

In the Pre-Fusion Power Operation Phase, operation of the ITER tokamak with the majority of thesystems required to achieve the Project’s goals will be demonstrated. This includes commissioningof the H&CD systems, fuelling systems, plasma control, Diagnostics, etc., which are required toprovide routine operation up to the nominal plasma current/toroidal field (15 MA/5.3 T) in L-mode
and up to (approximately) 7.5 MA/2.65 T in H-mode. These aims will be pursued by developing
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plasma scenarios that satisfy all the operational requirements in this phase, including those of coreedge plasma integration (ELM control, divertor power load control, etc.). In addition, developmentof control schemes and integration issues required to meet the needs of DT high-Q operation(including long-pulse/steady-state goals) will be investigated.
The PFPO Phase is divided into two sub-phases. In the first sub-phase (PFPO-1) a partial set of
Diagnostics, H&CD systems (20 MW ECRH and 10 MW ICRF – the latter is a working assumption
at present depending on whether the acceleration of 10 MW of ICRF is feasible or not), fuelling and
ELM control power supplies, will be installed for commissioning and operation. In the second subphase (PFPO-2) the H&CD, fuelling, ELM control systems etc. will be completed to their baseline
specifications.
The main objectives of PFPO-1 operation are:












Commissioning of the plasma control, interlock and safety systems with plasma to levels
supporting the experimental program (7.5 MA L-mode and 5 MA H-mode);
Establishing error field correction and the characterization of error fields without TBMs;
Commissioning of installed diagnostic systems, validation of data and integration into
control/ interlock systems;
Commissioning and establishing operation of divertor power load control schemes;
Commissioning and routine operation of the available ECRH and ICRF, if available,
systems to the installed power level for at least several tens of seconds;
Commissioning of disruption prediction, avoidance and disruption/runaway electron
mitigation systems and characterization of disruption loads;
Establishment of routine and robust plasma operation in L-mode to at least 7.5 MA;
Commissioning and exploitation of the Vertical Stability coils, Error Field Correction Coilsand ELM control coils (with an initial set of power supplies);
Exploration of H-mode operation in hydrogen and helium at 1.8T and investigation of ELM
control;
Initial physics characterization of ohmic, L-mode and H-mode plasmas;
Initial characterization of plasma-wall interactions in the ITER environment.

The main objectives of PFPO-2 operation are:








Commissioning and exploitation of the H&CD systems (HNB (33 MW), ICRF (20 MW),
ECRH (20 MW), DNB) in plasma operation to their installed power level for at least 50 s;
Commissioning of the additional Diagnostic systems, validation of data and integration into
control/ interlock systems;
Establishing routine error field correction with TBMs;
Commissioning advanced plasma control capabilities required to execute the experimental
program;Extension of the capability for disruption prediction, avoidance and mitigation up to
15 MA/5.3T L-mode plasmas
Exploration of the capabilities of the ECRH system to control NTMs;
Demonstration of divertor heat flux control capability in H-mode and L-mode to the highest
Paux = 73 MW;
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







Establishing H-mode plasma operation up to 7.5 MA/2.65 T (in H/He, as the power
threshold allows) and development of a reliable ELM control capability;
Demonstration of 15 MA/5.3 T L-mode plasmas in H;
Physics characterization of ohmic, L-mode and H-mode plasmas at the highest achievableparameters in terms of plasma current/magnetic field and input power in H/He plasmas;
First characterization of the current drive capabilities of the ITER systems and developmentof key ingredients of the long-pulse scenarios (e.g. target-q formation and sustainment for~10 s);
Time allowing, initial attempt at long-pulse (< 1000 s) H-mode operation at 7.5 MA/2.65 T;
Characterization of key aspects of plasma-wall interactions in the ITER environment (e.g.
fuel retention/removal, dust production, PFC erosion/redeposition).

Together with these high-level objectives, the operational experience and experimental results are
expected to provide key information regarding required upgrades to ITER systems, which might beneeded to support the achievement of the high-Q goals in DT. These are described in section 4.
Regarding plasma-boundary physics, the main high-level objectives of the PFPO phases are:
 Determine quantitative characteristics of heat loads in quasi-stationary plasmas and during
plasma transients (this strand of research would be closely linked to studies of disruption
heat loads and their mitigation);
 Characterize processes determining material erosion, transport and redeposition;
 Quantify hydrogenic retention and demonstrate techniques to be used later for control of
in-vessel tritium inventory (including need for routine wall conditioning/cleaning); Validate estimates of dust production and characterize dust generated; Validate the power handling performance of the Be first wall during limiter phases andcharacterize the power handling limits of the first wall during divertor operations; Characterize the operational domain of the tungsten divertor, including the establishment of
reliable methods for divertor power load control.
The PFPO phases will allow the characterization of key aspects of plasma-wall interactions (PWI)
in the ITER environment. Such measurements can begin as soon as plasma experiments are initiatedin PFPO-1, but the higher powers and longer pulse lengths which can be achieved in PFPO-2 willbe the first opportunity to access the new regime of PWI which ITER is expected to produce. In
addition, most of the diagnostics for dust/tritium retention/erosion will only be installed prior to
PFPO-2. Of primary importance is validation of the understanding of tritium retention, material
migration and dust formation to provide input on these issues before nuclear operations begin.
These analyses would also contribute to the refinement of the tritium/dust management strategy to
be implemented during the various phases of DT operation; it is understood that this strategy will befully validated during the DT phase.
Characterization of hydrogenic retention rates can be performed during both PFPO phases using gasbalance during hydrogen operations, assuming that the required hardware is in place. Given the
difference in the retention mechanisms of H isotopes and He in plasma-facing materials, gasbalance measurements during pure He discharges bring limited information, with the possible
exception of providing information on the background H level in the machine coming from the invessel PFCs. For as long as the beryllium first wall is in place, fuel retention in ITER is expected tobe driven mainly by co-deposition of hydrogenic species with Be. Gas balance measurements

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during PFPO-1 are made difficult by the low wall fluxes expected in this phase, given the low
available input power. Measurements in pure H would therefore likely provide only an upper
estimate of the hydrogenic retention and data for wall outgassing behaviour between discharges.
The tritium removal strategy in ITER foresees the use of thermal outgassing through the baking of
PFCs to 240C for the first wall and 350ºC for the divertor, respectively. In addition, conditioning
discharges (GDC, ICWC) can be used for fuel removal by ion-induced desorption and isotopic
exchange. The efficiency of these latter techniques should be studied early on, and, as for tritium
retention, it can only be evaluated during operations with hydrogenic fuels. A full bake is therefore
needed either during, or at the end of, PFPO-1 to provide a first indication of the wall release
behaviour.
The pulse duration, and resultant divertor fluence, during these early phases will however belimited, so that a full validation of the baking efficiency can only be obtained when sufficient codeposition has occurred in the machine, justifying the need for a dedicated campaign of repeateddischarges in PFPO-2, with trace deuterium if possible. On the contrary, the efficiency of ICWC forwall conditioning can be studied in detail during PFPO phases. One limitation is that isotopic
exchange of H and D can be performed only if D-trace experiments are possible, or if D-ICWC
discharges are permitted.
Another critical aspect of the plasma boundary physics during PFPO phases is the characterizationof PFC heat loads and the validation of models of power and energy deposition. Whilst the divertor
heat loads during the PFPO phases should be well within the design limits for any reasonable
divertor neutral pressure (Appendix C), wall heat loads may already reach their design limits during
the limiter start-up phase of diverted discharges depending, for example, on the degree of first wallalignment to the magnetic axis or misalignments between adjacent panels. Careful characterizationof the heat load pattern during limiter phases will therefore be required at the start of PFPO-1, andshould be performed at the main value of toroidal field chosen during the campaign. First wall heatloads, in particular for the upper panels close to the secondary X-point, should also be characterized
during divertor operations as a function of parameters such as primary-to-secondary X-point
distance, triangularity, density, power, confinement mode (L- and H-mode). The same exercise willneed to be repeated during PFPO-2 to help establish the limits on separatrix separation during FPO
and to check if upper main chamber power handling will be adequate for burning plasma
conditions.
Characterization of the near-SOL width in different operation regimes (L-mode, inter-ELM) will bea prime objective during PFPO phases to validate the divertor heat loads to be expected at higher
performance. Commissioning of heat load control methodologies must also be extensively
performed in the non-active operational phase since deuterium operation in the nuclear phase is
expected to rapidly achieve conditions in which divertor detachment control will be mandatory for
scenario development to high current and with tritium fuel. Early testing of power flux control,
particularly in H-mode, is also advantageous in the sense that stationary heat loads will be far fromtechnology limits and even uncontrolled ELM transient heat fluxes should be below damage
thresholds for the tungsten monoblocks. Such testing must, however, be mostly conducted in H
plasmas since detachment dynamics in He are very different and will not be representative of later
operation in D and T.
2.5.2

Assumptions for Pre-Fusion Power Operation

The assumptions made in developing the ITER Research Plan have been broken down into the four
operation stages to clarify what machine and system configuration is assumed to be available for60

ITR-18-003

each stage. The systems and subsystems required at each stage are described in [ITER_D_SNE6G8,
2016] and [ITER_D_TVG7YK, 2017]. In the following analysis, it is assumed that the First Plasma
campaign was successful in achieving full integrated commissioning of the operational ITER plantsystems, including a plasma of at least 100 kA for at least 100 ms duration at 2.65 T in H and/or He
and that the CS, PF, TF, and CC magnets were subsequently commissioned to their fully specified
operating voltage and current values without plasma. It is also assumed that the first 8 gyrotrons of
the ECRH system have been fully commissioned, at least for short pulse operation (> 100 ms
duration) at 170 GHz through one upper launcher. The PCS, CIS, and CSS have all been fullycommissioned for the functions required for First Plasma operation. Furthermore, it is assumed that
all commissioning of plant systems and diagnostics required for the subsequent operation phase thatcan be performed without plasma has been carried out successfully during the integrated
commissioning phase just prior to the restart of plasma operations.
It is planned to install the ELM and VS in-vessel coils prior to First Plasma, but it is assumed thatthey will be shorted with appropriate resistances to avoid arcing at the coil terminals and to avoid
large induced currents in the coils during First Plasma (FP) operation. Since the in-vessel coil
power supplies are not expected to be available for FP, full commissioning of these coils and theirpower supplies can only begin during the commissioning phase just prior to PFPO-1. It is assumed
that the VS in-vessel coil power supplies will be installed and ready to commission the coils before
the beryllium blanket modules are installed so that any issues found during coil commissioning canbe resolved before the coils are trapped below the blanket modules. To allow initial H-mode
operation during PFPO-1 and to begin to commission the ELM coils, it is assumed that at least nineof the ELM coil power supplies will be available and ready for commissioning of the ELM coils
before the blanket modules are installed. Although a temporary glow discharge cleaning (GDC)
system will be installed for FP, the permanent system will be installed and commissioned in
preparation for PFPO-1 plasma operation. Following commissioning of the in-vessel coils, theshielding blanket modules, including the beryllium first wall, and the divertor with tungsten PFCs
will be installed. The diagnostics required for the PFPO-1 phase (Table H-2 in Appendix H) will be
installed during Assembly Phase II. The remainder of the gas injection system (GIS) will be
installed through the lower ports, and one cask with two pellet injectors (PIS) will also be installed.
The disruption mitigation system (DMS) will also be installed during Assembly Phase II. The three
additional upper ECRH launchers and one equatorial launcher, together with the remaining 16
gyrotrons and their transmission lines will also be installed during Assembly Phase II to provide thefull 20 MW ECRH power during PFPO-1. For the purposes of the Research Plan development, it isfurther assumed that the first ICRF antenna system will be installed in advance of PFPO-1 toprovide an additional 10 MW of ICRF heating in PFPO-1, so that ICRF commissioning can begin
early in the experimental program with the aim of providing a total of 30 MW auxiliary heating
power for initial H-mode operation in this phase.
During Assembly Phase III, between the PFPO-1 and the PFPO-2 phases, a number of additional
systems will be installed and these will be integrated into the operational environment during
Integrated Commissioning III, prior to PFPO-2. This extended operational capability will include
the second ICRF antenna system (assuming that the first one is installed for PFPO-1 operations) toprovide a total of 20 MW ICRF heating, the diagnostic neutral beam (DNB) and two heating neutralbeam (HNB) systems to provide 33 MW neutral beam power, additional diagnostics, a further pellet
injector cask with 2 additional pellet injectors, and the first 6 Test Blanket Module (TBM) systems,
incorporating ferromagnetic structural materials. During PFPO-2, it is planned to commission thethree baseline auxiliary H&CD systems up to the full injected power capability of 73 MW. The
additional diagnostics to be installed for operation in the PFPO-2 phase are listed in (Table H-3 in61

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Appendix H). With a total of 4 pellet injectors, high fuelling capabilities to full performance will bepossible, as well as for ELM pellet pacing. The influence of the ferromagnetic TBMs on therequired error field corrections to avoid low density locked modes, as well as their impact on Hmode performance, will form one component of the experimental activity in PFPO-2.
2.5.3

Plasma scenarios for hydrogen and helium operation

During the PFPO phase of operation, plasma fuelling will be limited to hydrogen or helium, though,as discussed in section 2.5.5.12.2.1 and Appendix A, a limited experiment with low deuterium
concentration may be possible. It is expected that, whenever possible, preference will be given tohydrogen plasmas, since the overall fuelling, transport and recycling properties will be similar to
those of deuterium and tritium and since there is some remaining uncertainty about the impact ofhelium plasmas on plasma-facing materials. Nevertheless, given that the H-mode power threshold is
expected to be about 1.5-2.0 times lower in helium than in hydrogen, helium operation may be
unavoidable in order to obtain a high quality H-mode enabling ELM control and mitigation tests tobe undertaken during the PFPO phase. For these reasons, and given the uncertainties on the actual
H-mode threshold in ITER, scenarios have been designed for both hydrogen and helium plasmas.
In subsequent sections details of the most relevant plasma scenarios are presented, including steps
towards 15 MA operation, with an emphasis on H&CD conditions and on capabilities for longpulse operation. In view of several operational options that have been identified in the Research
Plan analysis following the development of the revised schedule within the Staged Approach, bothbaseline and optional scenarios are described.
The steps towards 15 MA operation described in the present document may be adjusted during theexperimental campaigns according to actual observations and achievements. These steps will be
progressive in magnetic field and plasma current, with some deviations from operation at 𝑞95 ~3 toallow the approach to the nominal 15 MA/5.3 T scenario to investigate carefully the issues and risks
associated with disruptions. Figure 2.5-1 shows an estimate of the operational range for each
H&CD system (EC, IC, NBI) for different values of magnetic field and plasma current. Due to theresonance conditions associated with the absorption of EC and IC waves, their operational ranges intoroidal magnetic field are constrained by the requirement that the power is absorbed sufficiently
close to the plasma centre. The operational range of the HNB system is constrained by theminimum plasma density required to limit shine-through power losses to an acceptable level. Steps
foreseen towards 15 MA operation are displayed as purple points on this figure.
The Staged Approach foresees the installation of the full ECRH system and a possible accelerationof the ICRF system installation, with one IC antenna installed in advance of the PFPO-1 campaign.
The full baseline H&CD capability with, in particular, the full power ICRF and HNB systems will
be installed in advance of PFPO-2.
Two principal areas on the installation of the H&CD systems required to meet the four-stage
approach remain the subject of R&D studies:The available EC frequencies and the associated pulse duration: to enable successful ECassisted breakdown and burn-through at third field (1.8 T), and a reliable path to H-mode forthe heating phase, second harmonic frequency (X2), EC assistance may be required.
However 170 GHz gyrotrons can only offer third harmonic (X3) EC assistance at this low
field. Hence, low frequency gyrotrons are under investigation, though potentially the output
power and pulse duration of the ECRH system will be more limited at the low frequencies
under consideration, as documented in Appendix G. Three options are foreseen for the 24
gyrotrons of the ECRH system:62

ITR-18-003

1) 24 single-frequency gyrotrons (SFG) at 170 GHz;
2) At least 16 SFGs at 170 GHz and up to 8 dual-frequency gyrotrons (DFG) at 104/170
GHz (UL);
3) At least 16 SFGs at 170 GHz and up to 8 SFGs at 110 GHz (UL).


The acceleration of one IC antenna (~10 MW) for the PFPO-1 campaign, for which apreliminary analysis is described in Appendix G.

Figure 2.5-1: Operational range of H&CD systems over the relevant BT, Ip and density range in ITER.
The EC, IC, NBI operation ranges are in blue, brown and green respectively.

All these possibilities should be envisaged in the description of the scenarios foreseen at each
operational step. The availability of H&CD systems is summarized in Table 2-3.
Scenarios in the non-active phase can be grouped in six main categories [Schneider, 2017-1]:







First plasma:
100 kA/2.65 TFirst divertor plasma: 3.2 MA/2.65 TFirst 𝑞95 = 3 plasma: 7.5 MA/2.65 TFirst H-mode plasma: 5.0 MA/1.8 T
Progressive steps:
7.5-9.5 MA/3.3 T,First 15 MA plasma: 15 MA/5.3 T

9.5-10.5-12.5 MA/4.5 T,

12.5 MA/5.3 T

For all these cases, it is likely that hydrogen plasmas will be developed first, followed by helium
plasmas where operation in helium is necessary, e.g. only if a good H-mode cannot be achieved in63ITR-18-003

hydrogen plasmas with the available H&CD power. Details of these scenarios are described in the
sub-sections below.
Table 2-3 – Availability and characteristics of H&CD systems in the four stages of the
Research Plan (tbc = to be confirmed by future R&D studies)StageFirst Plasma

PFPO-1PFPO-2

FPO-1

ECRH
systemOnly UL (8gyrotrons),
170 GHz,
6.7 MW

Full system, 170
GHz (+104/170 or
110 GHz, tbc),
20 MW

Full system, 170
GHz (+104/170 or
110 GHz, tbc),
20 MW

Full system, 170
GHz,
20 MW

ICRF
systemNone

1 antenna,
10 MW (tbc)

Full system,
2 antennas, 20 MW

Full system,
2 antennas, 20 MW
Full system,
2 injectors, 1 MeV
D beams, 33 MW
73 MW

HNB
systemNone

None

Full system,
2 injectors, 870 keV
H beams, 33 MW

Total

6.7 MW

28-30 MW (tbc)

73 MW

2.5.3.1 First Plasma
The first plasma will be attempted initially in hydrogen at a magnetic field of 2.65 T with a plasma
current of at least 100 kA for at least 100 ms, with a possibility to go up to ~0.5-1 MA for a few
seconds (~3s) for first plasma observations and first PCS operation in view of PFPO-1 preparation[Gribov, 2016]. A CS coil current of 20 kA/turn is planned as a target for magnet commissioning[Kavin, 2016-03] in preparation for attempting breakdown and burn-through [Kavin, 2016-11].
Complete burn-through could lead to a plasma current above 1 MA and steps will be taken to limit
the plasma current below this value since the first plasma protection components are designed tohandle disruptions up to 1 MA/2.65 T. In addition, currents approaching 1 MA may cause excessive
impurity generation from the steel limiter surface [Pitts, 2016].
H&CD capabilities:
ECRH system: only one upper launcher (UL), i.e. 8 gyrotrons delivering 6.7 MW at 170 GHz(X2).
2.5.3.2 First divertor plasma in PFPO-1
The first divertor plasma will be in hydrogen in L-mode at a magnetic field of 2.65 T with aplasma current of 3.2 MA. The goal is to establish initial divertor operation at the lowest possible
plasma current, with a flat-top duration of at least 10 s, as a target for the commissioning of various
systems, such as PCS, diagnostics, interlock systems, DMS [ITER_D_NSHUQ7, 2014]. The choiceof 3.2 MA is based on an analysis indicating that this is the minimum current allowing reliablemagnetic control with a full bore divertor configuration. The currents in the CS coils can be varied
from 12 kA/turn to 30 kA/turn, leading to a flat-top duration between 10 s and 200 s [Kavin,2015-08].
H&CD capabilities:
ECRH system: EL+UL launchers, 24 gyrotrons delivering a total EC power of 18-20 MW, withthe following possibilities:
64ITR-18-003

o Option 1: 24 gyrotrons delivering 20 MW at 170 GHz (X2).
o Options 2 and 3: at least 16 gyrotrons delivering 13.3 MW at 170 GHz (X2) and up to 8
gyrotrons delivering 6.7 MW at 104 or 110 GHz (O1) from the UL1.
2.5.3.3 First q95 = 3 plasma in PFPO-1 and PFPO-2
The first 𝑞95 = 3 scenarios will use hydrogen and helium plasmas at a magnetic field of 2.65 Twith a plasma current of 7.5 MA. At this combination of parameters, hydrogen plasmas are likely toremain in L-mode, while current expectations in the H-mode threshold scaling for helium plasmas
indicate that H-mode operation should be achieved with the H&CD power available in PFPO-2 andthere is some possibility of achieving helium H-mode operation at 2.65 T in PFPO-1 if sufficient
ECRH and ICRF power can be injected. The maximum flat-top duration will typically be of theorder of a few 100 s with a CS coil current varying from 30 kA/turn to 45 kA/turn [Kavin,2016-02; Kavin, 2016-03]. Operation above 40 kA/turn will consume the CS coil fatigue lifetime,and hence long-pulse operation will be addressed during the PFPO phase only if operational time
permits and long pulses of interest can be produced without significant consumption of the CS
fatigue life.
H&CD capabilities:


ECRH system:
PFPO-1 and PFPO-2: EL+UL launchers, 24 gyrotrons delivering a total EC power of 20 MWwith 170 GHz gyrotrons (X2) and possible 104 or 110 GHz gyrotrons (O1).

ICRF system:
PFPO-1: one antenna delivering 10 MW to the plasma.
PFPO-2: two antennas delivering 20 MW to the plasma.
Scenarios [Schneider, 2016]:
o Hydrogen plasmas: there is no good heating scheme. The best scheme is 2nd harmonic He4
heating (𝑓𝐼𝐶 = 40 𝑀𝐻𝑧) with an He4 concentration of ~20% leading to ~50% of single-passabsorption (SPA) dominantly on electrons, hence very broad.
o Helium plasmas: the best heating scheme is fundamental minority H heating (𝑓𝐼𝐶 =40 𝑀𝐻𝑧) with an H concentration of ~2-8% leading to a SPA of almost 100%.



HNB system:
PFPO-1: no NBI power.
PFPO-2: two injectors of 870 keV hydrogen beams delivering a total of 33 MW to the plasma;
however the actual power and energy that can be applied will be limited by shine-through power
levels, leading to an NB injected power varying from ~10 MW (in H plasmas), to ~20 MW (inHe plasmas) and to 33 MW at higher densities [Kim, 2016].

In order to compensate for the lack of a good ICRF heating scheme in hydrogen plasmas at 2.65 T,
one option is to heat small amount of He3 (<1%) at its fundamental resonance (𝑓𝐼𝐶 = 40 𝑀𝐻𝑧) at
3 T and/or 3.3 T in a H:He4 mixture with ~15% of He4 using the 3-ion heating scheme [Kazakov,
2015]. These scenarios could be applied both in PFPO-1 and PFPO-2. Under these conditions, the
SPA is estimated to be of the order of 80-90% shared between central electron heating and off-axis
ion heating on the high field side (HFS). These optional scenarios offer the possibility to reach Hmode in hydrogen plasmas when operating close to half field, thus enabling long-pulse operation1

Feasibility studies for SFGs at 110 GHz have not yet been performed and hence there is an uncertainty on the EC
capabilities in terms of power and pulse length at low frequency.

65

ITR-18-003

and ELM control at 𝑞95 = 3, and limiting the need for helium plasma operation at half field.
However, establishing a basis for the exploitation of this scenario in ITER requires further
experimental R&D to demonstrate efficient heating under analogous conditions to ITER plasmas. In
addition, adding 10-15% of He4 reduces the minimum allowable operational density with respect to
NBI shine-through. If observations from recent JET experiments [Hillesheim, 2016] are confirmed,
such a fuel mixture could also allow access to hydrogenic H-mode operation at lower input power.
However, W accumulation needs to be further investigated due to a possible increase of W
concentration with the dominant off-axis and reduced central heating. Furthermore, controlling He3
concentration in the presence of He4 may be challenging [Hellerman, 2017]. Finally, ion losses arestill to be assessed since they can be significant due to off-axis fast trapped ion acceleration andplasma response to 3-D magnetic fields (intrinsic and corrected by error field correction coils or toapplied fields for ELM control).
2.5.3.4 5 MA/1.8 T scenarios in PFPO-1 and PFPO-2
5 MA/1.8 T scenarios are designated as first H-mode plasmas in the context of the PFPO-1 phasesince they are aimed to enable H-mode operation in this phase with very limited auxiliary heating(between 20 and 30 MW depending on hypotheses on ECRH frequencies and ICRF accelerationplan). However these scenarios will be investigated both in PFPO-1 and PFPO-2, and both inhydrogen and helium plasmas. Maximum sustainable flat-top durations for these scenarios extend
from ~200 s to ~1400 s for CS coil current values varying from 20 kA/turn to 45 kA/turn [Kavin,2016-09]. Again, operation at CS coil currents above 40 kA/turn should be limited in order toreduce the consumption of the CS coil lifetime before the FPO phase. Operation at such low fieldleads to an overcompensation of the magnetic field ripple by ferromagnetic inserts, inducing aripple at the plasma separatrix of -1.3%, as described in detail in Appendix F. Ripple-induced fast
ion losses still need to be assessed in these scenario conditions. If this ripple level is found to be
critical, it can be reduced to -0.55% by increasing the plasma-wall distance [Kavin, 2017-01], but
doing so risks a significant degradation in the ICRF coupling performance.
H&CD capabilities:


ECRH system:
PFPO-1 and PFPO-2: EL+UL launchers, 24 gyrotrons delivering a total EC power of 20 MW.
For 1.8 T scenarios, the options 2 and 3 with up to 8 low frequency (104 or 110 GHz) gyrotrons
may be required to ensure X2 breakdown and burn-through. X2 absorption can be provided byeither 104 or 110 GHz gyrotrons; this also provides preheating of the plasma such that the X3absorption by 170 GHz gyrotrons becomes more efficient [Snipes, 2016-01]. This mode ofoperation is possible due to the fact that the absorption of the EC waves is central for both
frequencies at 1.8 T, leading to a synergetic effect, as shown in Figure G1-12.

ICRF system:
PFPO-1: one antenna delivering 10 MW to the plasma is assumed.
PFPO-2: two antennas delivering 20 MW to the plasma.
Scenarios:o Hydrogen plasmas: the best heating scheme is 2nd harmonic H majority (𝑓𝐼𝐶 = 53 𝑀𝐻𝑧)leading to a SPA of almost 100%.

2This needs to be confirmed by further modelling of ECRH scenarios with a combination of 104 or 110 GHz and
170 GHz frequencies at 5 MA / 1.8 T in a time-dependent transport simulation.

66

ITR-18-003

o Helium plasmas: the best heating scheme is 2nd harmonic H minority (𝑓𝐼𝐶 = 53 𝑀𝐻𝑧).
Preliminary analysis shows that this scheme leads to an SPA of almost 100%, which iscounter-intuitive, given that 2nd harmonic minority schemes are not expected to be very
efficient. The good absorption is associated with the fact that the absorbed power isinversely proportional to the square of the magnetic field, leading to a factor of ~ 10 increasein absorption at one-third field compared to that at full field.


HNB system:
PFPO-1: no NB power.
PFPO-2: two injectors of 870 keV hydrogen beams delivering a total of 33 MW to the plasma;
however, the total power and the injection energy that can be applied in such scenarios will belimited by the shine-through power level: injection should be possible at roughly 8 MW/500
keV in H plasmas and 16.5 MW/650 keV in He plasmas [Singh, 2017].

2.5.3.5 Progressive steps towards full current in PFPO-1 and PFPO-2
Steps in magnetic field and plasma current are illustrated in Figure 2.5-1 [Kim, 2016]. They have notbeen modelled in detail yet, and hence their description is still at a preliminary stage. These steps
are foreseen both in hydrogen and helium plasmas. They will likely be in L-mode since H&CD
capabilities are reduced at intermediate values of field and current, as shown in the operational
ranges of Figure 2.5-1.
Foreseen ECRH heating characteristics are given below, cf. Figure G-1:
 3.3T: At least 13.3 MW at 170 GHz (X3/X2) and up to 6.7 MW at 104 or 110 GHz(X2/X1).
 4.5T: At least 13.3 MW at 170 GHz (X2/X1), no resonance for 104 or 110 GHz.
 5.3T: At least 13.3 MW at 170 GHz (X1), no resonance for 104 or 110 GHz.
ICRF heating schemes are likely to use fundamental He3 minority in hydrogen plasmas (for an
ICRF frequency varying from 40 to 53 MHz for fields from 3 T to 5.3 T). In helium plasmas, thebest ICRF heating scheme is likely be fundamental H minority at 3.3 T and fundamental He 3
minority at 4.5 T and 5.3 T, though this is to be confirmed by further modelling.
The applicable NBI power and energy is directly correlated to the density and the possiblecombinations need to be accurately assessed for the different plasma currents/ densities foreseen inthese scenarios. The NB shine-through power depends on the fraction of the Greenwald density at
which it is chosen to operate, and this will be adjusted according to actual experimental conditions
and observations.
2.5.3.6 First full current plasma in PFPO-1 and PFPO-2
The first 15 MA/5.3 T (L-mode) plasmas will be developed in PFPO-2, since it is not considered
likely that there will be sufficient operational time available in PFPO-1 to develop the scenario andto commission reliable operation of the disruption detection and mitigation systems to this level ofplasma performance. However, performing full field/low current discharges, typically 3.5MA/5.3 T,
is foreseen during PFPO-1 in limiter configuration to develop disruption mitigation capabilities
with respect to runaway avoidance before applying the DMS at currents above 3.5 MA.
For the 15 MA/5.3 T scenario, the H-mode power threshold in hydrogen is expected to be wellbeyond the level of installed heating power, while it might just allow H-mode access in helium at
densities of around 50% of the Greenwald density. Nevertheless, the risks associated with operatingsuch high current plasmas just above the H-mode threshold are such that it is more likely that the67

ITR-18-003

development and demonstration of 15 MA plasmas during the PFPO-2 phase will be carried out inthe hydrogen L-mode regime. The flat-top duration can vary from ~10 s in ohmic plasmas to ~100 swith full auxiliary power (73 MW), for a CS coil current of 45 kA/turn.
H&CD capabilities:


ECRH system: EL+UL launchers, 24 gyrotrons delivering a total EC power of 20 MW, withfollowing possibilities:o Option 1: 24 gyrotrons delivering 20 MW at 170 GHz (O1)o Options 2 and 3: At least 16 gyrotrons delivering 13.3 MW at 170 GHz (O1); the up to 8
low frequency gyrotrons (104 or 110 GHz) have no resonance inside the plasma for fields
above 4 T.



ICRF system:
PFPO-1: one antenna delivering 10 MW to the plasma is assumed.
PFPO-2: two antennas delivering 20 MW to the plasma.
Scenarios [Schneider, 2016]:
o Hydrogen plasmas: the best heating scheme is fundamental He3 minority (with aconcentration <1%) in a H:He4 mixture (𝑓𝐼𝐶 = 53 𝑀𝐻𝑧) leading to a SPA of ~95%.
o Helium plasmas: the best heating scheme is also fundamental He3 minority (𝑓𝐼𝐶 =53 𝑀𝐻𝑧), with a He3 concentration of ~4-5%.



HNB system:
PFPO-1: no NB power.
PFPO-2: two injectors of 870 keV hydrogen beams delivering a total of 33 MW to the plasma.

2.5.4

Operations plan for PFPO-1

2.5.4.1 PFPO-1 Objectives
The PFPO-1 phase is the first plasma operations phase with the beryllium blanket and tungsten
divertor installed, allowing plasma currents greater than 1 MA and plasma durations beyond several
seconds. Initially, plasma initiation will need to be re-established in hydrogen at 2.65 T with the
new first wall conditions. This will be followed by the development of limited low elongation
plasmas at currents of up to 2 – 3 MA, together with the development of disruption and runaway
electron (RE) mitigation. Plasma magnetic control of the current, position, and shape, including
vertical stability control of elongated plasmas will lead up to 3.5 MA diverted plasma operation.
Routine operation will then be developed up to 7.5 MA to provide a suitable plasma configuration
for the commissioning of diagnostics and the ECRH power up to the full 20 MW and the first ICRF
antenna, which is assumed to be available for operation, up to 10 MW for at least 50 s duration.
Effective disruption and RE mitigation will have to be progressively established as the plasma
current is gradually increased. The toroidal field and plasma current will increase in steps beyond
7.5 MA/2.65 T up to 7.5 MA/5.3 T and, if time allows, possibly up to 10 MA. Time permitting, anddepending on the progress of the experimental program, the pulse duration may be extended to
beyond 100 s at lower plasma currents to qualify the H&CD systems for longer pulse operation in
PFPO-2.
A significant part of the initial operation program will be devoted to the commissioning of H&CD,
fuelling, control and diagnostic systems with plasma. For the last, in particular, the calibration anddetermination of the measurement capabilities of the installed diagnostics, as well as the resolutionof possible inconsistencies among the measurements of plasma parameters, are essential to allow anefficient progress of the experimental program. Table 2-4 lists the additional measurement
68ITR-18-003

requirements beyond those specified for FP established for the PFPO-1 campaign together with theadditional diagnostic systems which are planned to be installed in Assembly Phase II to meet theserequirements, when integrated with diagnostic systems already installed for FP. The requirements
focus on the need for basic characterization of essential plasma parameters, plasma control
capability and machine protection. There is typically more than one diagnostic system satisfying
each requirement, in part because the required parameter involves the integration of
Table 2-4 – Overview of additional measurement requirements and additional diagnostic
systems installed for the PFPO-1 campaign*
Measurement Requirement

Installed Diagnostic Systems

Magnetics for position, velocity, shape, MHD
mode structure and halo currents

•
•
•
•

Core profiles of ne and Te (TS, ECE)

• Core Plasma Thomson Scattering (1 laser)• Electron Cyclotron Emission• Toroidal Interferometer Polarimeter (TIP
measures line integrated density in EQ9)
• Density Interferometer Polarimeter (DIP
measures line integrated density in EQ8)

Ti and High-Z impurity content (XRCS andRadial Soft X-ray Cameras)• X-Ray Crystal Spectrometer• Radial X-Ray Camera••
Impurity identification and influxes, Zeff•
(visible and near UV spectroscopy, including•
Hα and visible bremsstrahlung, He, Be, Ne, Kr,•
Ar, W emission)••

High Frequency Sensors
Divertor Coils
Divertor & Blanket Rogowski Coils
Divertor Shunts

H Visible (partial)
Core Imaging X-ray Spectrometer
X-Ray Crystal Spectroscopy Edge
Visible Spectroscopy Reference System
Vacuum Ultra-Violet Edge
Divertor Impurity Monitor (partial)VUV Divertor

Radiated power distribution (Bolometry)• Bolometry System

Neutrons (Micro Fission Chambers, Neutron
Flux Monitors, Divertor Neutron Flux
Monitors, Neutron Activation System)**

•
•
•
•
•
•

Microfission Chambers
Neutron Flux Monitor Systems
Neutron Activation System
Divertor Neutron Flux Monitors
Neutron Calibration (2.5 MeV) (partial)
Neutron Facility Area

• Langmuir Probes
Divertor plasma characterization (incl. divertor• Pressure Gauges (cassette)
duct pressure and gas composition)• Residual Gas Analyzers (2 channels)
Visible/IR TV viewing (full coverage)• Visible/IR Equatorial Ports• Visible/IR Upper PortsPWI/Heat Load characterization• First Wall Samples• Divertor IR Thermography69

ITR-18-003• Thermocouples (Divertor)• Calorimetry (for testing with aux. heating)
Data for gas balance measurements• Pressure Gauges

In-vessel viewing system• In-vessel viewing system* The table lists the additional diagnostic systems planned for installation prior to PFPO-1 to meet the additional
measurement requirements (when integrated with the systems installed for FP). See Appendix H for a discussion of thescope of systems labelled as ‘temporary’ or ‘partial’.
** Neutron measurements are required to monitor neutron production during the non-active phase.

measurements from two (or more) diagnostics and in part to provide an element of redundancy (riskreduction) in the installed measurement capability.
Routine operation at 7.5 MA/2.65 T will be performed and is essential to support the experimental
program envisioned for this and later operation phases. This task includes qualification of theplasma shape and current control, development of robust scenarios for current ramp-up and rampdown, basic control of the plasma density, and robust plasma initiation. It requires that the first wall
protection and disruption protection have been successfully completed. It is expected that thisdevelopment will be performed in L-mode plasmas in H with He, a scenario required forcommissioning of the ICRF to highest powers (with the first harmonic hydrogen minority scheme).
He plasmas at 2.65 T will also be required to determine the scaling of the H-mode threshold withtoroidal field by comparing the requirements for 1.8 and 2.65 T plasmas.
Routine operation of the ECRH system at full power (note: this may not achieve 20 MW, dependingon the option chosen to provide breakdown/heating capability at 1.8 T) for several tens of seconds
is also necessary to support the physics program envisioned for this and later operation phases.
Operation at both 2.65 T and 5.3 T is expected. Proper aiming and localization must be verified. At
5.3 T, optimization of the polarization will be required.
Routine operation of the ICRF system at 10 MW for several tens of seconds, if available in thisphase, is necessary to support the physics program planned for this and later operation phases. This
includes integration of the PCS control of gas and antenna-plasma gap to optimize coupling,
qualification of the wall protection systems to handle antenna and fast particle loss issues, andensuring compatibility of the ICRF coupling with levels of non-axisymmetric fields from the ELM
control coils expected during ELM mitigation. While commissioning at low power can be
developed at 2.65 T in H plasmas, high power ICRF commissioning will require 2.65 T He plasmas(H minority first harmonic) and 1.8 T H plasmas (majority second harmonic).
Demonstration of a robust first wall protection system is required to progress beyond the basic
3.5 MA limiter operation. This includes interpretation of the signals from the protection cameras
and development of appropriate avoidance and response functions in the PCS, including termination
protocols.
Qualification of operation at 5.3 T is also necessary in the PFPO-1 phase, since the risk of runaway
generation is higher during full field operation, and this should be assessed at low plasma current assoon as possible. The commissioning of the ECRH system at 5.3 T is also important to assess therisk of stray radiation arising from reflection at the X-mode cut-off layer.
Routine disruption management must be established in this phase. The quantification of forces andrunaway generation and the effectiveness of the DMS to mitigate these should be performed as
early as possible to assess the need for upgrades to the DMS. Disruption management also includes

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development of the capability to predict conditions which are more likely to lead to disruption,
implementation of avoidance strategies, and a robust triggering algorithm for the DMS.
Assessment of intrinsic error fields and of the efficacy of the external correction coils to mitigate
their effects are essential in this operation phase. It is important to characterize the magnetic
geometry prior to the introduction of the TBMs in PFPO-2 to understand any effects that the TBMs
may introduce and to develop mitigation strategies.
It would also be desirable within the PFPO-1 phase to achieve:Plasma current operation above 7.5 MA, particularly to qualify disruption protection;
Plasma operation for > 100 s duration to qualify H&CD systems.

Operation above 7.5 MA is highly desirable in PFPO-1 to accelerate the approach to operation atfull technical parameters (15 MA/5.3 T) in PFPO-2. The DMS will be qualified for 7.5 MA/5.3 T
and it would be desirable to increase the plasma current up to 10 MA to demonstrate its
effectiveness in these conditions.
Operation of the ECRH and ICRF systems for > 100 s is important to test the in-vessel componentsof these systems as early as possible in the program to mitigate risk. Long-pulse operation will also
allow more effective use of machine time for the retention experiments if they are performed in thisphase (i.e. if the required hardware is installed).
The H-mode power threshold will be determined at 1.8 T (in H and He) and 2.5 T (He). If themeasured power threshold turns out to be consistent with predictions of the standard scaling (seeAppendix B), H-mode operation in 5 MA/1.8 T plasmas in H and He will be demonstrated; thesewill be the first ITER H-mode plasmas. Plasma parameters in these H-mode plasmas, uncontrolled
ELM losses, stationary divertor power loads, etc., will be characterized. ELM control will bedemonstrated with an initial set of power supplies for the ELM control coils and initial experimentson divertor power load control in H-modes will be carried out at this stage. These H-mode plasmas
will provide reference conditions in preparation for studies of the possible influence of TBMs on
H-mode performance and to develop mitigation schemes in PFPO-2.
In addition to the physics interests, it is expected that several control issues for H-mode plasmas
will be addressed during initial H-mode operation. These include shape control across the L-H and
H-L transitions, fuelling strategies for L-mode and H-mode, gap and gas control for maintaining
ICRF coupling, and early indication of the need for mitigation of impurity accumulation in theplasma centre. Addressing these in PFPO-1 would accelerate the PFPO-2 research programsignificantly.
The need to reduce the field to 1.8 T introduces a set of additional objectives:Commission the ECRH system at two frequencies (104 GHz/110GHz and 170 GHz), if
lower frequency gyrotrons are installed (options 2 and 3);
Develop plasma scenarios at 1.8 T;
Commission the ICRF system to heat at 1.8 T;
Commission the ELM control coils with the initial set of power supplies.

The option to operate at 1.8 T may require dedicated commissioning of the ECRH system (twofrequencies) and demonstration of its application to ITER scenarios. This may require the use of thelower frequency gyrotrons for plasma initiation and plasma preheating, and of the higher frequency
gyrotrons (170 GHz) for plasma heating in third harmonic. Similarly, the ICRF system would need
to be commissioned to allow H-mode operation in hydrogen plasmas at 1.8 T. It would be of71

ITR-18-003

particular interest to commissioning the ICRF antenna at the highest voltage stand-off capability inorder to allow a larger distance between plasma and antenna, since this may be required tominimize the TF ripple effects on these 1.8 T H-modes.
2.5.4.2 Establishing routine operation with plasma to 3.5 MA divertor configuration
PFPO-1 follows Assembly Phase II with the installation of the first wall panels and divertor. During
PFPO-1, the ITER Research Plan aims for a progressive commissioning of plasma operation andcontrol capabilities towards 7.5 MA at 2.65 T. For the initial phase of operation up to
3.5 MA/2.65 T in hydrogen, the intermediate milestones are:Commission robust plasma initiation;
Limiter operation at 2 MA, then at 3.5 MA;
Divertor operation at 3.5 MA.

The main commissioning items during this period are (i) the magnetic control, including verticalstability (VS), (2) documentation of disruption loads and effectiveness of mitigation methods (3)density control using gas and pellet fuelling, (4) the assessment of heat loads to the blanket anddivertor modules and (5) initial operation of ECRH and ICRF H&CD systems.
2.5.4.2.1 Commissioning of robust plasma initiation
PFPO-1 will start with commissioning and optimization of the plasma initiation and burn-through.
A robust operating space needs to be determined and the control of the plasma formation needs tobe commissioned. Overheating of in-vessel components or arcing due to stray EC radiation before
and during breakdown is potentially an issue and, if a robust breakdown/ burn-through strategy
without EC assist can be established, the routine use of ohmic breakdown would be the preferable
approach.
For operation up to 3.5 MA a precharge of the central solenoid (CS) of 20 kA is sufficient. The plan
envisages a start of operations at 2.65 T at maximum loop voltage using ohmic breakdown withprefill scans and PF coil bias scans for a first optimization of the timing of the plasma formation.
This will give the basis for the application of the EC breakdown assist, to obtain plasma formation(in case ohmic breakdown is not successful) and to expand the operation window for reliablebreakdown. Using EC assist requires that the injection of EC waves into vacuum is commissioned
due to high stray radiation (interlocks).
The operational domain for plasma formation should be documented in terms of gas dosing range(prefill), PF coil bias, timing of the gas and bias with respect to the loop voltage evolution and thetiming of the EC power waveform. Ideally the plasma formation should be commissioned at
different CS precharge currents (10 - 30 kA).
Ideally EC should only be applied with significant single pass absorption, i.e. electron temperatures
around 100 - 200 eV. Absorption may be judged observing the area of first reflection with an
infrared system comparing to a short pulse in the empty vessel.
Optional for this first phase are the commissioning of plasma initiation at 5.3T and using helium.
Both can be commissioned when required by the research program. Helium operation of ITER
could in principle also be achieved with plasma formation in hydrogen. This requires EC operationin O-mode. Experiments in present day devices show that break down assist is more effective inhelium and, therefore, faster with a shorter phase of high stray radiation. The effects during burnthrough are not too different in both gases.

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2.5.4.2.2 Limiter operation up to 2MA
The first demonstration of plasma magnetic control is planned at low currents of Ip ~ 2 MA and inlimiter configuration, with a flat-top phase of several seconds. For the shape measurement, thecomputed plasma boundary from magnetic data needs to be validated against other diagnostics suchas (infra-red) camera images of the blanket modules and probe measurements. Magnetic control
commissioning will go in several stages, from mainly preprogrammed plasma formation, Ip control,boundary flux control (or centroid control), tests of VS at low elongation (natural elongation,
~ 1.5) and verifying control gaps in limiter operation.
A first assessment of heat loads on the blanket first wall panels will be performed, althoughestimates of the PFC power loads during PFPO-1 indicate that during this period power loads on
PFCs should be well within their design limits. Additional time will be required for validating thewall protection techniques. Corrective actions need to be identified and tested on real pulses, in caselimiting conditions are reached, as part of the event handling development (see section 2.5.4.4).
Correction of error fields with toroidal mode number n = 1 will be performed by the external Error
Field Correction Coils. During the initial operation phase, algorithms for error field reduction will
be developed. The in-vessel (ELM control) coils could be used for an improved assessment andcorrection of the error field structure.
The DMS needs to be commissioned during early plasma operation, to start the documentation of
initial disruption forces (unmitigated, mitigated) and RE assessment in limiter plasmas as outlined
in section 2.5.4.6.1.
Initial technical commissioning and characterization of the fuelling system can be performed
without plasma. The commissioning of the plasma density control consists of two stages; the firstwith only the gas valves and a second stage with both the gas valves and pellet injectors. Bothbasic, simple control and model-based control need to be commissioned. Density control with gas
injection in limiter (2 MA) and different phases of the discharges such as the current rise and rampdown will be commissioned, initially with hydrogen gas fuelling alone. Density limits and fuelling
limits will need to be documented. The interferometer is essential for line-averaged electron density
measurements, associated with boundary reconstruction for the calculation of the average density.
For helium operation, planned later in PFPO-1, density control will need to be re-optimized and thedensity achievable in helium needs to be assessed.
2.5.4.2.3 Limiter operation up to 3.5 MA
After completion of the initial commissioning at 2 MA and assessment of the disruption forces,
operation will be expanded to 3.5 MA in limiter configuration with low natural elongation. VS
control will be commissioned with and without in-vessel VS coils, by gradually increasing theplasma elongation to nearly diverted plasmas. Tuning and optimization would include, if possible,
reduction of low frequency noise in the dz/dt diagnostic signal used in the feedback stabilizing loop.
As soon as plasmas with an electron temperature of several hundred eV are achieved, EC
commissioning can start. The principal issues to be checked are the beam polarization and thelauncher angles. Operation of EC at 2.65 T is planned, injecting X-mode and heating at the second
harmonic resonance. For the equatorial launcher (EL), in particular, this approach is potentiallyhampered by strong absorption at the 3rd harmonic (X3 heating) towards the plasma edge on the low
field side (LFS). Potential cross polarization (O-mode) may be strongly absorbed at the 2nd
harmonic resonance (O2-heating). These additional absorptions may complicate the verification ofpolarization and location of deposition. X3 and O2 heating significantly contribute at higher
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temperatures and densities. It may therefore be necessary to perform the polarization tests at thelowest currents for which a diverted plasma can be achieved. This needs further analysis using
beam tracing codes. It is noted already here that X3 heating can be avoided by increasing thetoroidal field for this intermediate operational point from 2.65 T to 2.75 T.
In parallel, a further assessment of the DMS system will be carried out for limiter operation up to
3.5 MA, together with the documentation of the disruption loads and the potential for creating RE(see section 2.5.4.6.1 for details).
Also heat loads to blanket modules, error field correction and density control may need further
assessment and optimization at 3.5 MA.
2.5.4.2.4 Divertor operation at 3.5 MA
It should be noted that operation in a divertor configuration at low plasma current (3.5MA) is
challenging from the point of view of control of the magnetic configuration. However disruptions at
low current have lower risk of localized damage to PFCs, allowing optimization of the plasma
magnetic control, optimization of the plasma-wall gaps, strike point control and stabilization ofplasma vertical displacements in the divertor configuration both with and without the VS in-vessel
coils. For divertor operation, additional validation with other diagnostics is required, e.g. to confirm
the location of strike points, plasma/wall separation etc.
First diverted plasmas at 3.5 MA will be at lower elongation ( ~ 1.5) to allow tests with andwithout VS coils. Operation will be extended to full bore at full elongation, once the control of theplasma boundary, the control of the strike points and VS have been commissioned.
A plasma scenario needs to be developed with current ramp-up, X-point formation andramp-down in divertor configuration using feedback control of plasma current, plasma-wall ‘gaps’
and strike point positions. In addition, scenarios with different rates of current ramp-up(with/without ECRH) and ramp-down will be explored. This will also allow validation/adaptationof the plasma transport models used in simulations describing these phases of the scenario.
Commissioning of event handling and ramp-down or shut down strategy/logic is important during
the initial divertor operation phase, laying the basis for routine divertor operation in ITER.
Commissioning of basic wall protection could be performed at much lower limits or by means ofartificially triggered events to commission the PCS response.
The density control needs to be tuned for divertor operation using both gas dosing and pellet
fuelling. Once pellets are actually being produced and injected, plasma operation will be required tovalidate the quality of the pellets entering the chamber. Tests will be necessary for pellets ofdifferent sizes.
Substantial commissioning of the DMS system will be performed at 3.5MA divertor operation, as
outlined in section 2.5.4.6.1. Commissioning of the H&CD systems (ECRH and ICRF) will also belaunched; ECRH having been restricted to plasma initiation and burn-through in the early phase ofoperation. ECRH will be used in the optimization of the plasma initiation and current rise, while
ICRF will initiate the first coupling studies (short pulse). Commissioning of the H&CD systems
with plasma is outlined in section 2.5.4.5.
2.5.4.2.5 Time required for obtaining routine divertor operation at 3.5 MA/2.65 T (hydrogen)During this first period of operation, in particular, the main activities relate to the commissioning ofsystems. Table 2-5 provides estimates of the time required to implement the main tasks at 2.65 T inhydrogen.
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Table 2-5 – Estimates of experimental time required for initial activities at 2.65 T
Deliverable

Plasmainitiation

2 MA limiter
3.5 MA limiter
3.5 MA
divertor

Description
At 2.65 T, hydrogen: optimize gas, PF coil bias, andtiming; EC power scan: start with ohmic breakdown, andthen add EC; document robust operational range; optional:
optimization for 5.3 T or helium.
The first demonstration of plasma magnetic control
together with an assessment of heat loads to first wall; error
field correction.
Optimization of VS control at higher elongation;
check gap control.
Optimize shape control to full bore plasma operation;
optimize current rise and current ramp-down; commission
event handling; commission basic wall protection.

Timerequired(days)
7

11
11
28
57

Density control
and fuelling
1st DMS tests

HeatingParasitic to the main deliverables
Optimize density control in limiter configuration using gas
dosing; commission density control for divertor plasmas
including pellets.
Disruption studies in limiter plasmas at 2 MA and 3.5 MA,
including RE studies; disruption forces and their mitigationin divertor plasmas at 3.5 MA; document disruption forces(mitigated and unmitigated).
ECRH for plasma initiation, short pulses (<1 s) of ECRH
in limiter and divertor plasmas; start of ICRF coupling andvoltage stand-off studies.

(9)*(15)**(15)*** Included in time required for limiter and divertor commissioning.
** Detailed (and counted) in other sections of the IRP.

2.5.4.3 Validation of diagnostic data and demonstration of measurement consistency
As discussed in section 2.4, diagnostic calibration and validation activities will be initiated during
Integrated Commissioning (e.g. magnetic measurements, neutral pressure measurements etc.).
Subsequently, following FP, but before the blanket modules are installed, extensive measurements
will be made of the (toroidal) magnetic axis of the toroidal field to align the blanket modules (seesection 2.4.3), which will already require validation and demonstration of measurement consistencyof the magnetic field measurements with modelling of the field in the absence of eddy currents.
With FP, there will also be the opportunity to validate and demonstrate the measurement
consistency between the magnetic diagnostics and modelling, including eddy currents in the
conducting structures. From early in PFPO-1, as each of the diagnostics is commissioned, theirmeasurements will need to be validated and the consistency between various measurements willneed to be demonstrated, particularly for control and investment protection functions. Thisvalidation and consistency process will continue throughout all the operational phases as theinstalled diagnostic capability expands.

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Once routine operation is underway, the process for the data validation will not require dedicatedplasma time for the majority of the cases. Detailed commissioning and validation plans for the
diagnostic systems are not yet fully developed, but it is expected that many systems can becommissioned, and their data validated and cross-checked, under the standard plasma scenarios
which will be operated as the experimental program develops.
Nonetheless, dedicated pulses for diagnostic commissioning are included in the planned run-time
estimates and progress to the target plasma scenario will be performed in steps after diagnostic
validation. For example, plasma boundary and equilibrium reconstructions will be used to check theconsistency of magnetic measurements from different poloidal and toroidal locations once theplasma current exceeds several MA, to ensure adequate signal-to-noise ratios in the magnetic
measurements. Plasma shape and position control require verification that the accuracy of theplasma boundary calculation is within the expected uncertainties. Plasmas with sufficiently large
plasma-wall gaps for limiter plasmas will be established before operation of configurations with thefull bore reference shape is undertaken. Similarly, before achieving an acceptable plasma control,
the full equilibrium reconstruction is required to allow isoflux control. Transitions from limiter todiverted plasma will then be designed to guarantee the protection of the in-vessel components
against uncertainties in plasma shape.
After achieving sufficient levels of plasma current and density, comparisons of Thomson scatteringand interferometer density measurements, for example, will be used to cross-calibrate the various
density measurements and ensure consistent densities, including synthetic diagnostic modelling tosimulate the expected signals within the estimated inversion errors of chord-averaged measurements
compared to local measurements. Similarly, comparisons of kinetic pressure derived frommeasurements of electron and ion temperatures and densities will be made with magnetic
measurements of the plasma stored energy to check diagnostic consistency. With the beginning ofplasma heating, measurements of the ion temperature and of the current profile will assume
increasing importance, though the measurement capability in this area will be limited (seeAppendix H, section H.3) until PFPO-2, when the HNB, DNB, Charge Exchange Recombination
Spectroscopy (CXRS) and Motional Stark Effect (MSE) systems will be operational. Diagnostic
cross-calibrations and absolute calibration certainly will also require some time.
Overall, diagnostics required for investment protection will also have to be rigorously validated andchecked for consistency across multiple measurements. Plasmas with reduced risk for investment
protection will be used until accurate data validation and consistency have been obtained. For
example, the plasma shape will be cross-checked with visible camera images that will provide thecontact point during the limiter phase and the strike-point locations during diverted operation.
During the PFPO-1 phase, as additional diagnostics come on line and plasma conditions anddurations improve, diagnostic data validation will work towards applying automatic procedures tocompare different measurements of the same parameter, using data from different locations orderived from different methods of measuring the same or similar parameters, or using comparisonsof synthetic diagnostic models. Automatic routines will check measurement consistency among
different diagnostic signals and provide firstly off-line, and eventually real-time, validation of
signals within the plasma control system simulation platform (PCSSP), to allow the replacement ofsignals that are inconsistent and to ensure valid signals, particularly for plasma control andinvestment protection functions.
More sophisticated models of the plasma behaviour will also be used within the Integrated
Modelling and Analysis Suite (IMAS) to check measurement consistency with the expectations
from modelling. Any discrepancies will be used to help validate measurements from different76

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diagnostics or to improve synthetic diagnostic modules and theoretical models. As more diagnostics
are brought into full operation and increasing levels of additional heating power are commissioned
throughout the PFPO-1 phase, it will be possible to check the consistency of more sophisticatedmeasurements and the associated models of plasma behaviour, including particle and energy
transport models, whose validation and consistency are required both to ensure reliable operation inlater operational phases and to provide a detailed understanding of physics processes in burning
plasmas. This activity will therefore be essential preparation for subsequent high plasma
performance operation and fusion power experiments.
Based on experience in current devices, 10 run days have been allocated in PFPO-1 for dedicateddiagnostic commissioning, but, as noted above, the diagnostic commissioning and validation
activities will be considerably more extensive than implied by this figure.
2.5.4.4 Commission control, interlock and safety systems
2.5.4.4.1 Plasma Control (Conventional control)
During initial operation in PFPO-1, all of the basic control functions will be commissioned as wellas commissioning/verification of the PCS supervisory monitoring functions, event handling logic,
and the interaction with the CIS. All technical commissioning that can be carried out without
plasma for PCS control of the actuators, the communications among plant systems, supervisorysystems, and the PCS failure monitoring will be performed before plasma operation.
The strategy for PCS commissioning in PFPO-1 will be aligned with the overall logic of the
Research Plan, consisting of three stages with two options for further development:
1) commissioning of magnetic control up to robust diverted operation at 3.5 MA/2.65 T;
2) commissioning of magnetic control up to 5 MA/2.65 T;
3) commissioning of magnetic control up to 7.5 MA/2.65 T;
further development option 1) H-mode operation at 5 MA/1.8 T to commission initial ELM control
techniques; and option 2) commissioning of magnetic control up to ~10 MA/5.3 T, initiating thedevelopment towards 15 MA L-mode operation, which depends on progress in the development of
disruption and runaway mitigation. This phase will first commission initial plasma current, positionand shape control with limited plasmas, followed by diverted plasma operation and thecommissioning of reliable vertical stability control at high elongation. Equilibrium reconstruction
will be required in the first phase as well as error field control, particularly to correct for n = 1 error
fields. Density control will be commissioned for both gas and pellet fuelling. ECRH breakdown
assist will need to be recommissioned with the beryllium first wall and divertor, and ECRH control
will also need to be commissioned for burn-through, heating during the plasma current flattop andduring the ramp-down. ICRF commissioning of the first antenna will include plasma position,
shape, and localized edge density control to optimize ICRF coupling. Further commissioning willbring the ECRH power up to 20 MW and the ICRF up to 10 MW. If the H-mode threshold is within
the available power of ≤ 30 MW, first H-mode operation may be developed at 5 MA/1.8 T,
allowing first tests of ELM control with pellet pacing and with a subset of the ELM control power
supplies. Initial forecasting and event handling schemes will be developed to anticipate
controllability boundaries and to allow evasive actions, to optimize plasma operation and avoid
disruptions. Depending on the development of disruption and runaway electron mitigation
techniques, magnetic control commissioning could be extended up to ~10 MA at 5.3 T toward theend of the PFPO-1 campaign.
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Magnetic control
Control functions:






Plasma initiation with feedback control on the prefill neutral pressure;
Current, shape, and position control (including current saturation avoidance techniques);
Vertical stability control (with and without in-vessel coils);
Plasma boundary control through L-H and H-L transitions for H-mode operation;
Commissioning of error field correction with the dedicated error field correction coils.

A primary objective will be to develop the capability to properly control the shape of the plasma
through all the plasma pulse phases including transitions and to guarantee the required plasma-wall
gap clearances.
At the outset, real-time plasma boundary reconstruction and the vertical stabilization system will becommissioned and validated using real-time plasma simulation techniques in PCSSP so that theiroperational commissioning can begin with the pulses foreseen from the FP campaign. It will be
prudent to use larger plasma-wall clearances and a substantial margin on vertical stability until thenecessary control techniques can be fully validated and optimized.
Density control
Control functions:Gas injection system commissioning including calibration and validation of the models usedin control simulations;
Pellet injection technical commissioning with plasma (availability of a measurement of thepellet entering the vessel is not confirmed at the time of writing this document);
Validation of the feedback control loop: multiple scenarios by means of single to multiple
valves; the use of the pellet injection for density control should be commissioned as well
(reduced configuration of the system);
Validation of the density control schemes by using different diagnostics inputs.

H&CD system control
Starting with technical commissioning:Technical commissioning with plasma (i.e. 20 MW EC power injected for a short time);
ECRH technical commissioning: commissioning and validation at low power of the controlof the local deposition;
ICRF (one antenna): technical control for power, antenna phasing and frequency sweeping;
initial power coupling control development; IC for wall conditioning.

Initially, the ECRH control system will be tested without plasma, i.e. launcher setting, polarization
setting, polarization adaption as launchers move. This should be tested with a low power source upstream of the polarizers. This also includes an experimental verification of the assumed propagationof the polarizations through the network of waveguide switches. The interaction of PCS with thelower-level EC-launcher and polarizer control can (in principle, though not necessarily in practice)also be tested with the vessel open (safety measures around the moving launchers should be
implemented). Testing power switching by the PCS may be possible into dummy loads.
With plasma, polarization shall be optimized by minimizing the non-absorbed power as detected by
the infrared view of the respective HFS area. This may have to happen already at low plasma
current. The deposition is verified by power modulation and comparison to Fourier-transformed78

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Electron Cyclotron Emission (ECE) measurements. If X3-heating is significant, this is the point todecide if the toroidal field shall be raised by 0.1 T or if the equatorial launcher shall be used in Omode at 2.65 T (weak Electron Cyclotron Current Drive (ECCD)). Interlocks will have to be tested
with plasma; these may include interlocks to impose lower limits on plasma current, density,
electron temperature (in case of X3 or O2 heating). Stray radiation monitoring with a large numberof ECRH sniffer probes at various poloidal and toroidal locations is planned. The in-vessel elementsof the ITER ECRH system will be, in many respects, similar to those of the new W7-X system,
which has started operation (10 MW, 2000 s). It is strongly recommended to take advantage of theexperiments that will be undertaken at W7-X during the next five years to gain experience for
ITER. A potential safety issue relates to possible damage to the vacuum window, which requires apriori testing of the respective safety system (close fast valve, switch off beam).
Exception handling
Requires:Commissioning of the event handling system logic (this phase requires
commissioning/validation each time a new logic implemented);
Commissioning of the algorithms/data processing for event detections (i.e. vertical stability
margin for VDE avoidance, island tracking technique and diagnostics required later for
NTM suppression, etc.).

Note that each new exception handling event and algorithm will require commissioning to
implement it in the system.
Real-time recognition of off-normal events, such as the development of a significant population of
runaway electrons or approaching the predicted boundary for vertical stability, and the automatedinitiation of counter measures, including discharge termination, should be validated and introduced
into operation as soon as practical. Real-time protection for critical PFCs will need to becommissioned as a prerequisite for sustaining reliable operation in the next phase in which thepower and energy input will increase. These functions will be included in the plasma control system
functionality but an additional level of protection will be included in the central interlock system, toprotect against plasma events, with a high level of integrity.
Since the ITER Plasma Control System (PCS) will have an unprecedented level of sophistication, it
is clear that commissioning of its more complex functions will continue in parallel with plasma
operation throughout much of the experimental program. The PFPO-1 phase will lay the foundation
for an extensive program in plasma control, complementing progress in the science program as therequired level of fusion plasma performance is developed to achieve ITER’s goals.
2.5.4.4.2 Commissioning of the interlock system
The investment protection system is a collection of dedicated Plant Interlock Systems (PIS) and aCentral Interlock System (CIS) to protect against exceeding operational limits, to ensure that thesystems and components are protected against self-induced faults, faults generated by other systems
and unintended consequences of plasma operation.
The following steps will be taken to validate the interlock system:Technical commissioning (communication);
Commissioning of the interlock trigger from each of the plant systems;
Commissioning of the diagnostics used for operations (i.e. cameras);Commissioning of the PCS-CIS interface;79ITR-18-003



Commissioning of the avoidance/mitigation activities in coordination with the PCS (i.e.
plasma stop and termination by means of CIS trigger);Commissioning of the dedicated diagnostics including links with the CIS.

The operation of DMS via CIS for PFPO-1 is discussed in detail in section 2.5.4.6.
While the PCS will act as the first line of defence for investment protection, the CIS will always
back up the PCS for ultimate investment protection even for complex functions during plasma
operation, including first wall and divertor heat load protection. With each step up in plasma current
and heating power, at least one day of run-time dedicated to commissioning first wall and divertor
protection is included in the estimated run-time.
2.5.4.4.3 Safety System commissioning
For ITER, a set of use cases has been identified where the plasma is either involved in or affectedby safety functions. It is then mandatory to demonstrate that the safety events are detected and thecorresponding actions are performed. At present the only action associated with a Safety event
during operation is the activation of the Fusion Power Shutdown System (FPSS). The system has tobe tested eventually during dedicated plasma operation, probably with reduced performance to
avoid the risk of damage to internal components. Additionally all the critical event detection has tobe validated and tested similarly to the approach expected for the investment protection functions.
At present the only Protection-related (i.e. safety-related) plasma parameter is the plasma current.
No formal decision has been made on the functionality of this parameter but commissioning andvalidation for the plasma current measurement is expected as well as for the associated safetysystem function.
2.5.4.4.4 Control commissioning deliverables
For PFPO-1, the control commissioning deliverables include:















Plasma initiation with the beryllium first wall and tungsten divertor at 2.65 T in hydrogen;including neutral prefill feedback and ECRH breakdown control;
Initial CIS plasma operations related functions;Plasma current, position, and shape control for limited plasmas to 3 MA;Real-time equilibrium reconstruction;
ECRH control for burn-through and heating during the plasma current flat-top and rampdown;
Plasma current, position, and shape control for 3.5 MA/2.65 T diverted plasma;
Vertical stability control to 3.5 MA/2.65 T;
Density control with gas and pellet injection;
Error field control, particularly to correct for n=1 error fields;
Further commission CIS plasma operations related functions;Plasma current, position, and shape control up to 5 MA/2.65 T;
ECRH operation at powers of up to 20 MW;
First ICRF antenna, including plasma position, shape, and localized edge density control for
ICRF coupling;
ICRF operation at powers of up to 10 MW.

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2.5.4.5 Heating & Current Drive (H&CD) commissioning (ECRH, ICRF) in plasmas up to
7.5 MA/2.65 T
The four-stage approach foresees the installation of the full ECRH system and a possible
acceleration of the ICRF system implying the installation of one IC antenna, in advance of the
PFPO-1 campaign, presently under study. The full ICRF system, the HNB and DNB will be
installed in advance of PFPO-2. The H&CD commissioning schedule is directly linked to this plan.
Each H&CD system requires a commissioning period without and with the plasma. The present
analysis focusses, for the most part, on the H&CD commissioning with the plasma, together withthe risk assessments and necessary diagnostics and control.
ECRH commissioning:
The early ECRH commissioning has to be undertaken prior to FP and is described in 2.4. The
PFPO-1 phase will benefit from the full ECRH power capabilities, i.e. the set of 24 gyrotrons
delivering a total EC power of 20 MW. As described in Appendix G, three options are considered:
1) 24 single-frequency gyrotrons at 170 GHz; 2) At least 16 single-frequency gyrotrons at 170 GHz
and up to 8 dual-frequency gyrotrons at 104-170 GHz; 3) At least 16 single-frequency gyrotrons at
170 GHz and up to 8 single-frequency gyrotrons at 110 GHz. The two last options are foreseen to
enable second harmonic assisted breakdown (X2) and plasma pre-heating in case of operation at
1.8T. At this stage where these options are still under investigation, the three of them have to beconsidered in the commissioning plans. All gyrotrons are assumed to be conditioned up to the torus
windows ahead of the operation with plasma. The ECRH commissioning should demonstrate thefollowing capabilities:

Demonstration of local deposition (correct polarization, separately for each gyrotron andfrequency) with an appropriate launcher control. If the deposition initially occurs at an
unexpected location, reasons may be manifold (launcher angles, X3 parasitic absorption, ray
tracing parameters, equilibrium, mapping routines, analysis of ECE data, …).



Launching 20 MW at short pulse via the equatorial launcher (EL) and/or the upper launchers(ULs).
Short pulse conditioning of all beam lines and parallel operation of all gyrotrons (screeningof interference of magnetic fields) and power supplies shall be demonstrated, along with thefull power capacity of launchers. It is important to use all beams of one launcher sinceinternal stray radiation adds up. These launchers will not have operated with more than one
beam at a time before they are installed. Sufficient time needs to be allocated for thisconditioning.

The effect of X3 heating at 2.65T for the equatorial launcher shall be checked. If crucial, themagnetic field may need to be increased by 2-4% or the EL polarization to be changed fromX- to O-Mode.

The X1 reflection for O1 heating scheme at 5.3 T has to be proven to be lower than 1%.
If the 1st harmonic is used for heating (high fields in ITER) then O-mode is required. Any
cross-polarized fraction will be reflected back at the X-Mode cut-off focussed on the LFSvessel-wall. ITER assumes 1% cross polarization corresponding roughly to a power densityof 1-10 MWm-2 at the location of the first reflection on the outer wall. Such an optimistic
estimate is the basis for the window design. Windows are typical components endangeredby incompletely absorbed EC-beams. ITER has to demonstrate here that these design values
for cross polarization can be matched (or that no window is hit by one of the reflected81

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beams). A proper detection of the absolute amount of cross polarization needs to be
developed. In present day devices cross polarization has been demonstrated to be below 3 %close to the resolution limit. Since ITER will use stray radiation bolometers, it may bepossible to sweep the reflection toroidally across such a device ramping the density in theplasma, but this scheme needs further assessment.
Long-pulse heating above 50 s shall be achieved. At this stage, the whole ECRH system andthe ITER vessel are cooled. The cooling is designed on the basis of predefined heat-fluxes.
On thin metal surfaces directly on top of the cooling channels (beam scrapers, etc.) no
problems are assumed, since those loads are well defined. There are usually regions moredistant from the next cooling channel for which a certain temperature gradient is allowed toremove a typical average energy input. Any anisotropy of the surface heating of a
substantial fraction of 1 MWm-2, which is the average neutron energy absorption in the wall,
may increase these gradients on long time scales finally leading to hot spots. These possible
issues need to be detected and ameliorated before the nuclear phase. Spill-over losses due toa large asymmetric mode content also need to be addressed (5% are allowed, which allows
for example beam shapes with 2 maxima).

As stated above, for the PFPO-1 campaign the full ECRH system will be commissioned with theplasma, i.e. the 24 gyrotrons, delivering a maximum power of 20 MW, using both the ULs and EL.
This commissioning duration is expected to be around 75 days. It will mostly be carried out in
7.5 MA/2.65 T scenarios, with 2nd harmonic X-mode ECRH absorption, requiring a low density andmodest temperature (n ~ 1019 m-3 and T ~ 1 keV) to achieve full absorption of the ECRH power.
This commissioning period does not need to be exclusive and could be associated with other plasma
operations. The ECRH commissioning plan will consist of the following steps:







Progressive increase of the ECRH power from 1 to 20 MW;
Progressive increase in the pulse length from 10 ms to 10 s;
Commissioning of the beam polarization sweeping by steering the mirrors to their fullrange; this requires around 5 sweeps per beam (8 beams for each of the 4 UL and 24 beams
for the EL), requiring 3 s for one mirror sweep – the operational constraint is the necessityof constant plasma parameters during a 3 s sweeping period;
Launchers are slowly conditioned to long pulses, with a growing factor of 1.4 for the pulse
duration, up to a duration of ~ 50 s; multiple launcher commissioning is possible in thesame pulse;Power modulation capability needs to be tested during this commissioning period.

Further ECRH commissioning time will probably be necessary as the toroidal field is raised
towards 5.3 T, for which the heating scheme would be fundamental O-mode absorption. As theplasma scenarios are developed towards higher current and field, it will be necessary to switch fromthe 2nd harmonic X-mode heating scheme to the fundamental O-mode scheme. In addition, theoption for operating at a toroidal field of 1.8 T would require an allocation of further
commissioning time to the ECRH system.
The ECRH commissioning requires the measurement of electron temperature profiles, and density
profiles, the latter associated with real-time ray-tracing modelling needed to predict the wave
propagation properties during the discharge and, if necessary, to trigger a controlled shutdown of
the ECRH to avoid wall damage. Clearly magnetic reconstruction is also needed – at some point an
accurate determination of the radial localization of the principal rational q-surfaces will also be
required to deal with the growth of NTMs (a more detailed assessment of the stage of theexperimental program at which NTMs are likely to appear is necessary).
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Risks associated with ECRH system commissioning are mostly due to stray radiation emitted
during breakdown and burn-through phases. Errors in the polarization setting and localized density
fluctuations at the separatrix (e.g. due to ELMs) could induce extra radiation losses within the mainchamber, possibly leading to some PFC damage.
ICRF commissioning:
Feasibility studies on accelerating the ICRF system to provide an antenna delivering 10 MW ofpower during PFPO-1 are underway, as described in Appendix G. Nevertheless, the present
assumption for the development of the IRP (agreed by ITER management) is that one ICRF antenna
will be operational in PFPO-1 and two antennas will be operational for PFPO-2. The
commissioning of the ICRF system must therefore begin during PFPO-1. Further commissioning is,of course, required during PFPO-2 in order to test the cross-talk of the two antennas and to adjust
the matching for H-mode operation.
Some ICRF commissioning will be carried out in vacuum prior plasma operation; antennaconditioning at high voltage in vacuum is necessary: (i) to confirm the operational strategy foravoidance of multipactoring conditions and (ii) to condition the system progressively to high RF
voltage, using trains of short pulses of increasing voltage, in order to establish the reliablemaximum operating voltage, for all foreseen operating frequencies.
The start of the commissioning of the ICRF system with the plasma will take place at half-field.
Part of this commissioning does not need a particularly efficient ICRF heating scheme, hence it canbe started in hydrogen and carried on in helium plasmas3. The commissioning period with theplasma is estimated to be around 2 to 4 weeks, and can be supplemented by commissioning during
plasmas targeted at other goals. The commissioning of the ICRF system will be mostly carried out
in L-mode, but extra commissioning time will be required in H-mode plasmas (hence mostly in
PFPO-2), as described in section 2.5.5.2. The commissioning can be split into two different
categories, early and advanced commissioning. Early ICRF commissioning is discussed here whileadvanced commissioning is split between this section and section 2.5.5.2, since some of theactivities imply the presence of the two antennas and L-H transition capabilities.
ICRF early commissioning (requires 5 to 20 s pulse dedicated time):






Assessment of the antenna coupling and behaviour of the automatic matching system onplasma completing the commissioning of the relevant algorithms with CODAC, at low
power;
Dedicated tests of the arc detection system, an essential component of the ICRF system
protections, to be validated both in L- and H-mode plasmas;
Test of the common antenna phasing, i.e. dipole, current drive and 00ππ phasing;
Test of the real-time frequency sweeping capability (± 1 MHz);
Test of the power modulation capabilities.

ICRF advanced commissioning (requires ~ 60 s pulse dedicated time and thus will be addressed as
the development of long pulses allows):
Progressive increase of the coupled power;

3

In H plasmas at 2.65 T, the envisaged ICRF heating scheme relies on direct electron heating by the fast magnetosonicwave, for which wave absorption is less efficient than with the minority heating schemes available in other plasmas orat higher magnetic fields. In He plasmas at 2.65 T, the envisaged ICRF scenario is fundamental minority H heating,which is an efficient heating scheme.

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





Progressive increase in pulse length: test of long-pulse operation at high power;
Demonstration of RF coupling control via gas puffing and control of the gap between theseparatrix and the antenna in L-mode;
Assessment of the production of impurities by ICRF;
Assessment of the compatibility of impurity gas puffing, used for divertor protection, with
the ICRF coupling performance in L-mode;
Test of the effect of 3-D fields (ELM coils) on the RF coupling and heat loads; first walland divertor protection with ICRF heating, i.e. localized loads due to fast particles, must be
verified.

Additional commissioning time could be required for 1.8 T scenarios where the ICRF heating
scheme would be 2nd harmonic fundamental H heating (primarily in H plasmas, but possibly also in
He plasmas, although the efficiency of ICRF schemes in He plasmas at 1.8 T has not yet been fully
assessed).
Several risks are associated with this commissioning. There is a possible antenna multipactoring
issue that could pose a risk to the ICRF system. Uncertainties on RF-sheaths for a given strap
configuration are large and this effect could lead to further localized wall erosion damage via
impurity acceleration. Uncertainties on the SOL density and on the distance between the separatrix
and the antenna lead to uncertainties on the RF coupling performance. Successful commissioning
will require the monitoring of the antenna front face by the visible and infrared cameras, andavailability of the RF arc-detection system. Finally, control of the minority ion concentration will
be desirable to adequately implement the desired heating scenarios. The overall operational range ofthe H&CD systems is summarized in Figure 2.5-1 where low toroidal magnetic field (1.8 T)
scenarios are also included.
2.5.4.6 Disruption management program in PFPO-1
This section describes the experimental program required with respect to disruption prediction,
avoidance, mitigation and loads. It is important to note that the program elements described here arestrongly interlinked with each other and with other elements of the Research Plan. Especially,
scenario development and the gradual increase in plasma current will have to go hand in hand withthe disruption alleviation activities listed here. Capabilities for disruption forecasting and avoidance
and their commissioning in PFPO-1 will be covered in section 2.5.4.8.
2.5.4.6.1 Disruption Mitigation System (DMS) commissioning/optimization
The present design of the DMS results in a multi-dimensional parameter space for the mitigation
scheme. Multiple injectors in several locations with different injection quantities and species
compositions for thermal load mitigation and for runaway electron suppression will be available[Baylor, 2015; Maruyama, 2015]. It is expected that by the time the commissioning work withplasma will start, the parameter space will be narrowed down with the experience gained frompresent devices. Nevertheless, it is highly likely that achieving all disruption mitigation goals will
only be possible with fine adjustments of the injection parameters during ITER operations andtherefore a substantial number of dedicated pulses to disruption studies will be required in PFPO-1
and PFPO-2. Also, the risk involved if mitigation is not complete during the optimization activities
requires a careful and thus pulse intensive approach. Note that due to uncertainties in the DMS
design and the most optimal injector set, the final configuration of the injectors may still evolve
over the next few years.

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Commissioning and optimization of the DMS aims at: (a) Thermal load mitigation, (b) Electromagnetic (EM) load mitigation, and (c) RE avoidance validation. The commissioning work in
PFPO-1 will start at low current and 2.65 T to reduce risks during the initial phase of DMS usage.
The commissioning program will be strongly interlinked with the gradual increase in plasma current
when developing plasma scenarios. For validation of runaway suppression it will also be important
to run at toroidal field of 5.3T and low plasma current early in PFPO-1.
(a) Thermal load mitigation
Thermal loads are caused by both the plasma thermal energy loss during the thermal quench and by
conductive losses of the magnetic energy in weakly radiating current quenches. In both cases, themitigation strategy aims at radiating a large fraction of these energies. Whereas the high-Z
quantities required to radiate the magnetic energy on the timescale of the current quench is expected
to be low and compatible with EM load limits, the number of particles required for the thermal
quench is uncertain and might be in conflict with allowable current decay times (see also 2.5.4.6.2).
In both cases, the quantities have to be validated and the injection scheme has to be adapted if
required. The compatibility of the thermal load mitigation scheme with the runaway avoidance
scheme is mandatory. The margin to runaway formation has to be kept large enough while
optimizing thermal load mitigation.
The following Milestones have to be achieved in PFPO-1:
2.5.4.6.1.1 Confirmation of current quench heat load mitigation for 7.5 MA operation
Prerequisite to run 7.5 MA scenarios.
Experimental strategy:Confirm mitigation in target plasmas of varying magnetic energy, starting at low plasma
currents;
Confirm required quantities and injection scheme to achieve 100% radiated power fraction
during the CQ;
Validate compatibility of heat load mitigation with RE avoidance scheme.

2.5.4.6.1.2 Confirmation of thermal quench heat load mitigation in L-mode
Prerequisite to run at thermal energies above melt thresholds.
Experimental strategy:Confirm mitigation in target plasmas of varying thermal energy in L-mode;
Confirm required quantities and injection scheme to achieve high radiation fractions during
the thermal quench;
Validate compatibility of heat load mitigation with RE avoidance scheme and EM load
mitigation scheme.

(b) EM load mitigation
EM load mitigation aims at reducing the halo current amplitude and asymmetry during VDEs while
keeping the current quench slow enough to not exceed limits on eddy current forces on in-vessel
components. Model predictions have to be validated to allow projecting requirements towards
higher energies.
The following Milestones have to be achieved in PFPO-1:
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2.5.4.6.1.3 Confirmation of EM load mitigation scheme for 7.5MA operation
Prerequisite to run 7.5 MA scenarios.
Experimental strategy:Confirm mitigation in target plasmas with varying magnetic energy vertically stable and indeliberate VDEs;
Confirm dependency of current quench rates on injected quantities, species and on theinjection scheme;
Validate halo current mitigation goals;
Validate compatibility with RE avoidance scheme.

The DMS commissioning activities, as well as the validation of EM loads during disruptions, will
require operation at 5.3 T as early as possible and preferably in the early phase of PFPO-1, during
limiter operation with currents up to 3.5 MA and during x-point operation with up to 7.5 MA.
(c) RE avoidance validation
Of high priority during DMS commissioning is to establish a mass injection scheme that guarantees
the suppression of runaways to acceptable limits. Due to the severe energy loads that can be caused
in case of substantial runaway formation, the initial tests of the operational space of the DMS withrespect to runaway generation have to be performed at currents of the order of 2-3 MA. The limiter
configuration and a high Bt will provide the most favourable conditions for runaway formation andare thus the worst case testbed [Reux, 2015]. As extrapolation from these conditions to higher
currents has non-negligible uncertainties, further tests to validate the mitigation scheme with respect
to runaways have to be done at higher currents. These tests will require proceeding in small current
steps, which has been taken into account in the required number of pulses for DMS commissioning.
The following Milestones have to be achieved in PFPO-1:
2.5.4.6.1.4 Confirmation of the injection scheme with maximum margin to RE formation
Prerequisite for using the DMS in closed-loop operation.
Experimental strategy:Gradual increase of Bt to 5.3T at 3.5 MA in limiter configuration;
Variation of injection species and injection scheme towards thresholds for RE formation.

2.5.4.6.1.5 Confirmation of the margin to RE generation in diverted configuration
Prerequisite to use DMS in closed-loop operation above 3.5 MA in divertor configuration.
Experimental strategy:Target plasma of 3.5 MA at 2.65 T in divertor configuration;
Variation of injection species and injection scheme towards thresholds for RE formation
starting from confirmed scheme for limiter configuration.

The estimated number of pulses/days required for DMS commissioning/optimization in
PFPO-1 is 560/43.

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2.5.4.6.2 Disruption Load Validation
Validation of the heat loads and their scaling with plasma parameters is essential to define therequirements for mitigation in terms of mitigation efficiency, but also in terms of the prediction
success rates and false alarm rates of the disruption prediction and detection system which will bepart of the PCS with the aim of triggering the DMS. Establishing scalings of VV and BM
electromagnetic loads is a prerequisite in order to go to higher plasma currents within the
progressive start-up approach. Disruption models will help to achieve a high accuracy towards
higher currents and to reduce the number of experimental data points needed to establish thescaling. Model validation will be therefore an integral part of this work. These models will beespecially important if thermal and electromagnetic loads make it necessary to limit the maximum
current for unmitigated disruptions in order to avoid unacceptable damage to the first wall or thedivertor and to avoid the occurrence of category III electro-magnetic loads on the VV.
The following Milestones have to be achieved in PFPO-1:
2.5.4.6.2.1 Confirmation of EM load scaling models
Prerequisite to increase current beyond 7.5 MA.
Experimental strategy:Variation of plasma current at 2.65T over a range sufficient for projecting to 15 MA, but
considering possible melt damage at higher current levels;
Measurements of halo currents and forces/displacements on components including the VV
during deliberate VDEs;
Confirm predicted asymmetry and dynamic amplification;
Confirm impact of plasma parameters e.g. internal inductance, elongation, current.

2.5.4.6.2.2 Confirmation of melt thresholds for thermal loads during the current quench
Prerequisite to increase the plasma current above the predicted critical current of about 6 MA.
Experimental strategy:Monitoring of first wall heat fluxes during upward and downward deliberate VDEs with
gradually increasing plasma current.

2.5.4.6.2.3 Confirmation of melt thresholds for the divertor
Prerequisite to increase the thermal energy beyond the predicted critical energy of 25 MJ.
Experimental strategy:Deliberate major disruptions and downward VDEs with gradually increasing thermalenergy.

2.5.4.6.2.4 Confirmation of melt thresholds for the first wall
Prerequisite to increase the thermal energy beyond the predicted critical energy of 50 MJ.
Experimental strategy:Deliberate upward and downward VDEs with gradually increasing thermal energy;
Possibly conduct post-thermal quench mitigation if higher currents are required to prevent
excessive current quench heat loads.
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The estimated number of pulses/days required for Disruption Load Validation in PFPO-1 is130/10.
2.5.4.6.3 Disruption Mitigation System (DMS) trigger generation
Achievement of the target disruption rates requires the development of avoidance and prevention
strategies, whilst the ability to attain the target mitigation rates depends on the development of
reliable prediction schemes. In early operation, disruption prediction will start with basic thresholdbased indicators such as mode amplitudes. As operation moves to more complex scenarios, thepredictor will have to be adapted and more input signals will need to be added. As mentioned
above, the requirements on the mitigation success rate for the loads during the current quench are
extremely demanding when 15 MA operation is approached. These rates will be achieved by adding
also indicators that a disruption is ongoing, such as the vertical plasma position or the current spikeat the thermal quench [Pautasso, 2016]. Statistical methods such as SVM require training, but
present developments are focussing on reducing the size of the training set and to improve
portability. Such schemes can complement threshold-based approaches. Approaches using physicsbased models including real-time modelling and automated disruption event characterization arepresently under development to further improve disruption prediction capability while minimizing
false positive predictions for these new schemes. Such schemes will be fully-developed by PFPO-1
and will be exploited.
The disruption predictor within the PCS is a module that solely aims on deciding to trigger the
DMS. Different injection schemes may be chosen to optimize the mitigation action for various
disruption situations. The PCS will also be responsible to decide about the injection scheme basedon plasma parameters, plant system status and DMS status. Note that the Central Interlock System(CIS) can, upon input from plant systems or from the plasma current measurement, also initiate a
mitigated disruption by triggering the DMS. In that case the injection sequence will either be thelatest issued by PCS or a default sequence. Testing and defining the injection sequences are part of
the DMS commissioning work.
Targets for disruption rates and mitigation success rates have to be defined as mentioned in 2.3
when taking steps in plasma current or thermal energy. However, confirming that these targets areachieved before taking the next step in performance cannot be done by a statistical approach since
this would require too many pulses or disruptions. Therefore, it will be mandatory to assess theeffectiveness of the chosen prediction algorithms by dedicated tests accompanied by modelling
validation. Within the threshold-based approach, it is expected that the database from present
experiments allows knowing the relevant parameters to consider in the threshold tests and to narrow
down threshold values based on validated scalings, physics-based modelling, and event
characterization analysis. Sufficient operational time has been planned to establish and test thealgorithms in ITER.
The following Milestones have to be achieved in PFPO-1:
2.5.4.6.3.1 Disruption detection ready for high current operation
Prerequisite to increase current beyond the cat. III threshold. The required reliability depends onthe expected loads and might vary for different plasma currents (c.f. sections 2.3 and 2.5.4.6.2).
Experimental strategy:Establish algorithms for detecting the thermal quench, the start of the current quench and a
vertical displacement;88ITR-18-003



Test the algorithms during low current operation and deliberate disruptions, if necessary.

2.5.4.6.3.2 Thermal quench prediction sufficient for L-mode operation
Prerequisite to run high energy L-mode beyond the melt threshold. The required reliability depends
on the expected loads and might vary for different thermal energies (c.f. sections 2.3 and 2.5.4.6.2).
Experimental strategy:






Establish threshold-based algorithm for predicting the thermal quench;
Implement statistical methods and physics model-based methods in parallel;
Perform dedicated disruptions to optimize the predictor if necessary;
Implement and test decision schemes in the PCS for the injection sequence;
Test the algorithms during low current operation and deliberate disruptions, if necessary.

2.5.4.6.3.3 Injection sequences ready for high current operation
Prerequisite to increase current beyond the category III threshold.
Experimental strategy:

Establish injection sequences for early injection before the thermal quench (TQ) and for late
injection into the CQ;
Verify RE avoidance and current quench (CQ) mitigation efficiency.

2.5.4.6.3.4 Injection sequences ready for elevated thermal energies
Prerequisite to run high energy L-mode beyond the melt threshold.
Experimental strategy:

Establish injection sequences for pre-TQ injection;
Verify RE avoidance, TQ and CQ mitigation efficiency.

The estimated number of pulses/days required to establish the DMS trigger function in
PFPO-1 is 135/11.
2.5.4.7 Development of reliable operation at 7.5 MA/2.65 T
2.5.4.7.1 Establishing 7.5 MA/2.65 T divertor plasmas with heating
Following the development of routine diverted plasma operation up to 3.5 MA and 2.65 T and thecommissioning of disruption and RE mitigation, diagnostics, plasma control, interlock, and safetysystems and initial commissioning of the ECRH and ICRF H&CD systems at relatively modest
plasma current levels, plasma scenarios will be developed up to 7.5 MA at 2.65 T. The plasma
scenarios will be initially developed in hydrogen, but a significant amount of operation in helium
will also be required to commission and develop ICRF heating schemes, which may allow H-mode
operation if the H-mode threshold is within the available power limit.
The first step will be to increase the plasma current to 4 MA at 2.65 T and further develop plasma
current, shape, and position control in hydrogen. This is a small increase in current so that thedevelopment should be straightforward to quickly ensure robust magnetic and density control aswell as improved disruption and RE mitigation. Depending on the level of error fields found due tocoil manufacturing and installation tolerances, error field correction may also require further

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development. Although it is a modest increase in current, routine operation at 4 MA is considered
the next deliverable milestone.
The development of 5 MA, 2.65 T discharges will be a significant increase in plasma current thatwill approach the upper limits for which melt damage from unmitigated disruptions is not expected.
Thus, at this stage, care will have to be taken assessing first wall heat loads following disruptions toensure investment protection. Further development of plasma current, position, and shape control
will be required including robust vertical stability control. Full commissioning of the ECRH andICRF heating systems at this current level in helium plasmas would help to assess disruption heatloads at still relatively low plasma current. Plasma termination scenarios for both standard rampdown and for emergency terminations will need to be developed at this stage to prepare for higher
current operation. Routine operation of 5 MA plasmas is the next deliverable milestone.
A further increase in plasma current to 6 MA will require more development of plasma current,
position, and shape control emphasizing robust vertical stability control at higher plasma current tominimize possible VDEs. Melt damage from unmitigated disruptions is expected at this current
level and there is increased risk of damage from RE. Thus, particular attention will need to be paid
to develop robust disruption and RE mitigation techniques. Further development of termination
scenarios controlling impurity accumulation and vertical stability in the ramp-down will also be
required. Routine operation of 6 MA plasmas is the next deliverable milestone.
Depending on the observed scaling of disruption forces and heat loads, it may be feasible to thendevelop 7.5 MA, 2.65 T scenarios, which will be the first scenarios at q95 = 3; care will be requiredin the development as such plasmas are expected to be more susceptible to disruption-inducing
MHD phenomena. Plasma current, position, and shape control emphasizing robust vertical stability
control will require further development. Core fuelling with pellet injection may be required at thiscurrent level. Central electron heating with ECRH and ICRF should be developed to control
impurity accumulation in the core. Sawtooth control should also be developed as large sawteeth are
expected at low q. Additional first wall and initial divertor heat flux control should also be
developed at this stage. Advanced control commissioning that will include capabilities for
disruption avoidance in PFPO-1 will be discussed further in section 2.5.4.8. Further development ofplasma termination scenarios is required to avoid impurity accumulation and ensure verticalstability in the ramp-down. Routine operation of 7.5 MA, 2.65 T plasmas is the next deliverable
milestone. The pulse duration of these scenarios should then be extended out to ~50 s in helium toprovide a standard ITER scenario for development of the heating schemes. Since this scenario is thehalf current, half field equivalent of the baseline 15 MA, 5.3 T scenario, development of robust
operation at 7.5 MA, 2.65 T is required to commission all systems in PFPO-1, particularly if the Hmode threshold were low enough in helium plasmas to fully develop H-mode operation and ELM
control at q95 = 3.
2.5.4.7.2 Studies of off-axis ECRH at 7.5 MA/2.65 T
As discussed later in section 2.5.5.11, the plan to expand L-mode operation from 7.5 MA/2.65T to
15 MA/5.3T in hydrogen plasmas in PFPO-2 relies on several intermediate Ip/Bt steps with q95 =3-4. This is also the basis for the development of L- and H-mode plasmas in FPO, as described in
sections 2.6.3.4, 2.6.3.5 and 2.6.4. For several of the scenarios in the development sequence it will
not be possible to ensure that the ECRH power will be deposited centrally and, for some extreme
cases in this step ladder, ECRH ~ 0.5 (peak of the power deposition profile). This raises the question
as to whether W accumulation may occur in such L-mode plasmas or not, thus potentially90

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increasing disruptivity. If this is the case, some steps in the initially proposed step-ladder will beomitted at the cost of larger changes of Ip/Bt between consecutive steps.
Due to the significant implications of this issue on the further development of scenarios in the IRP,
an assessment is included in PFPO-1. In general the problem of W accumulation with off-axis
heating is found to be more serious for H-mode conditions (including exit phases) than in stationary
L-modes in present experiments, given the reduced edge transport in the former. It is, therefore,possible that by ensuring that the W divertor influx remains low in ITER (low sputtering and highprompt redeposition as discussed in Appendix C) W accumulation may not occur in L-mode
plasmas even when ECRH heating is not fully on-axis.
The assessment will consist of a series of L-mode discharges at 7.5MA/2.65T with ECRH heating
at a level of ~ 14-20 MW and a range of plasma densities (<ne> = 0.2 – 0.8nGW) with pellet
fuelling and gas fuelling. In these conditions the injection angle of the ECRH equatorial launcher
will be adjusted from providing dominant central heating (ECRH ~ 0.2) to off-axis heating ECRH ~0.5 and W density and divertor plasma parameters/W influxes will be measured. The aims of theexperiments are:a) to determine which is the level of off-axis ECRH heating that, if exceeded, causes W
accumulation, andb) to determine if this level depends on divertor plasma conditions (influenced by the separatrix
density) and/or core density profiles (modified by pellet fuelling compared to gas fuelling).
The total number of experimental days estimated to be required for this assessment is 3.
2.5.4.8 Advanced control commissioning in PFPO-1
Several control systems will need to be commissioned to support ITER physics research in the three
primary research phases of the Staged Approach. These systems should aim to fulfil allcommissioning tasks well before their intended use, and preferably one operational phase earlier
than required to allow full system checkout under normal ITER operations. This will best reveal thepractical requirements of each system beyond what is envisioned and planned for prior to their use
in operation.
Once routine plasma operation is achieved and robust basic magnetic and kinetic control schemes
allow standard plasma pulses to be easily reproduced, more advanced control functions can becommissioned. Advanced control here refers to control functions that go beyond basic plasma
current, position, and shape control, beyond basic density control, and beyond basic heating system
control. The more advanced control functions include kinetic control functions such as: first walland divertor protection,  and stored energy control, rotation profile control, and minority species
control; MHD and error field control functions such as: ELM, NTM, dynamic error field and
RWM, and sawtooth control; and supervisory control functions including exception handling andmore advanced actuator management functions. The physics requirements for all control functionsof the PCS are described in the PCS Physics Requirements Document for the Preliminary Design[ITER_D_TVUZP2, 2016], and the detailed interface requirements between the PCS and each of
the plant systems and diagnostics can be found in the PBS-47 interface sheets that are listed in the
PCS Document Production Plan [ITER_D_RD59AJ, 2016]. Note that all disruption related control
in PFPO-1 is described in section 2.5.4.5. The run-time estimates given here are for commissioningof advanced control schemes, but further development of the physics R&D needed to develop
robust control is likely to require significantly more run-time as part of the physics program.

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2.5.4.8.1 Plasma kinetic control
First wall protection is one of the first advanced control functions that will need to becommissioned before plasma heat loads on the first wall are sufficient to exceed the power handlingcapability of the beryllium first wall panels. The first wall shaping has been designed based on theheat flux e-folding length presently scaled to ITER from current experiments to withstand steadystate limiter plasma operation up to about 7.5 MA with additional ECRH power in MW ≤ theplasma current level in MA (see Appendix C for more details). Because of uncertainties in theextrapolation of that scaling to ITER, however, it is prudent to avoid limiter plasma operation
beyond about 3.5 MA, when diverted plasmas should be achievable. A few discharges with special
shapes with plasma currents between 3.5 and 7.5 MA will be needed to fully commission first wallheat flux control, to investigate heat loads from plasma contact at different points along the firstwall, in particular to compare the protection of the high heat flux panels and the normal heat flux
panels. An estimate of 10 days of commissioning discharges for first wall heat flux control are
expected to be required to ensure adequate control as the plasma current is gradually increased.
Note that if the plasma current is raised beyond 7.5 MA in PFPO-1, robust first wall protection
must be demonstrated in both the PCS and the CIS to ensure investment protection of the first wall.
Divertor protection will not be required in the PFPO-1 phase since there will be at most 30 MW of
auxiliary heating power and the divertor is not expected to be overloaded for auxiliary heatinglevels under 40 MW (see Appendix C for more details). However, it will be important to begin todevelop divertor protection during the PFPO-1 phase before it becomes essential for investment
protection. Initial tests of divertor heat flux control algorithms can begin once routine diverted
plasma operation is achieved. With the limited auxiliary heating power available in PFPO-1, it may
not be possible to fully demonstrate effective divertor heat flux control, but it may still be possible
to control the limited heat flux reaching the divertor target. An estimate of 5 days commissioning is
expected to develop initial divertor heat flux control in PFPO-1.
Once the ECRH and ICRF systems are nearly fully commissioned, initial tests of β and plasma
stored energy control will be possible. While these control schemes are not required during
PFPO-1, development of effective β and stored energy control should begin in this phase to prepare
for PFPO-2. Before plasma parameters approach q95 = 3 at 7.5 MA, 2.65 T, it will be valuable todevelop β and stored energy control to better control MHD stability, particularly to achieve longer
pulse durations at nearly full auxiliary heating power. This can be attained by relatively simple
proportional-integral-derivative (PID) control algorithms, or more sophisticated physics modelbased controllers [Goumiri, 2017]. Various sensors could be used for this real-time control. The
simplest and less accurate approach would be to use a diamagnetic flux measurement and asimplified-yet-established equilibrium model to compute the stored energy in real-time. A moreaccurate approach would be to utilize the computed plasma stored energy produced by the real-timeequilibrium control in the PCS. The timescale of interest for plasma stored energy or  is theplasma energy confinement time, E, which is expected to be far longer than the control system
latency. The E can be occasionally altered in a non-predictable way by plasma physics phenomena
such as microinstabilities, which make such control necessary to maintain steady-state conditions.
Control of the plasma stored energy or  has already been performed in many tokamak devices, andsuch control is well understood. An estimate of 2 days commissioning is allocated to begin todevelop β and stored energy control in PFPO-1.
Another advanced control technique worth attempting during the PFPO-1 phase is rotation profile
control. While intrinsic plasma rotation is likely to be difficult to control, there may be some
limited control with ECRH or ICRF heating. In addition, even with the limited set of available
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ELM control coil power supplies in the PFPO-1 phase, the application of 3-D fields from the ELM
control coils could also be attempted for rotation profile control as observed in some experiments[Garofalo, 2008; Sabbagh, 2016]. While rotation profile control is not required in the PFPO-1
phase, initial attempts to develop rotation profile control in this phase will provide valuable
opportunities to assess different techniques that will become more important in later operational
phases to stabilize MHD instabilities and improve plasma performance. Rotation profile
measurements of both the core and edge should be available from X-ray crystal spectroscopy
systems 55.ED and 55.EI.
For more than a decade, the physical phenomenon of neoclassical toroidal viscosity (NTV) [Shaing,
2015] has been established as a technique to alter the plasma rotation in a tokamak in a controlled
fashion when a largely non-resonant, weak three-dimensional (3-D) magnetic field is applied [Zhu,2006]. The effect has been used routinely in tokamak devices primarily to slow the plasma rotationby providing drag. However, the effect can also accelerate the plasma from near-zero values,
originally found to provide rotation in the direction counter to the plasma current [Garofalo, 2008].
More recent experiments have examined a generalization of NTV physics and have demonstrated
plasma acceleration in the same direction as the plasma current as well, while producing large,
favourable plasma rotation shear and rotation speeds far exceeding ITER simulations of plasma
rotation in the pedestal region [Sabbagh, 2016]. In this case, the plasma velocity is altered toward
the so-called NTV offset rotation value, which changes with plasma minor radial position.
This 3-D field actuator can then be used either under open or closed-loop control to alter the plasma
rotation. Note that the initial toroidal plasma rotation profile at any time before the application of
the 3-D field will not be zero, but rather the so-called, but well-established ‘intrinsic rotation’ of theplasma, which depends on several plasma parameters but for which a first-principles plasma
physics model is not yet available. However, the NTV effect can effectively slow the local value ofthe plasma rotation when it is larger than the NTV offset rotation, and speed up the rotation when it
is slower than the NTV offset rotation value. The effect of slowing the plasma rotation using a
predominantly non-resonant 3-D magnetic field spectrum is robust and well-established. The abilityof increasing the plasma velocity by NTV has also been demonstrated in tokamaks, but with lessexperimental experience. Present experiments (2016+) are being conducted to gain further
experience with the generalized NTV offset rotation effect. Note that while the NTV offset rotation
effect has been shown to produce plasma rotation well above ITER simulations in the pedestal
region, based on present research it is expected that the effect will be limited to producing rotation
values near plasma diamagnetic drift values, which may still be adequate for ITER needs.
The key timescale for rotation control active feedback is the plasma momentum diffusion time,
which is far slower than the expected system latency. Real-time measurement of the plasma rotation
will be required if closed-loop feedback is desired. Note that open-loop control might be sufficientfor plasma operation during PFPO-1. An estimate of 2 run days has been allocated to initial tests ofrotation profile control in the PFPO-1 phase.
Another more advanced control that should be commissioned in the PFPO-1 phase is the control ofminority species for ICRF heating schemes and the mixture ratio of different plasma fuelling
species. To commission ICRF heating from the first antenna, the control of minority species density
within a majority background fuelling species will be required in PFPO-1. The diagnostics
available in PFPO-1 that can at least provide some information on the minority content and species
mixture include the visible H system (55.E2) and the vacuum ultraviolet system (55.E3). The
commissioning of minority species control will be included in the overall run day allocation for
ICRF commissioning. With only hydrogen and helium fuel available in PFPO-1, fuel mixture
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control may not be required in this phase, except possibly for some H-mode experiments. Still, it
would be valuable to begin to develop fuel mixture control experiments even in this phase as part ofchangeover experiments between hydrogen and helium.
2.5.4.8.2 Magnetohydrodynamics (MHD) and Error Field control
With a limited set of ELM coil power supplies and both the ECRH and ICRF H&CD systems
commissioned in PFPO-1, it will be possible to begin to develop advanced control schemes forseveral MHD instabilities and for error field correction. Depending on the level of error fields
found in ITER due to coil manufacturing and installation tolerances, it may already be necessary tocorrect for error fields with the external correction coils at low plasma current to avoid low density
locked modes in the current rise [Buttery, 2000; Luxon, 2003]. The correction of small, nonaxisymmetric magnetic fields in a tokamak generated by the real three-dimensional, non-toroidallycontinuous nature of device components is important to maximize plasma performance and yieldsuperior plasma stability. These ‘error fields’ can change in time throughout the plasma pulse
duration. The plasma itself can amplify such fields, and such amplification can itself be dynamic innature (time varying in amplitude and toroidal phase). Typically, the most deleterious error fields
have toroidal mode number n = 1, with varied poloidal field spectra.
The correction of device-generated error fields, called ‘dynamic error field correction’ (DEFC), or
those caused by plasma amplification of these fields, or due to global plasma modes, such as theRWM (whether it be stable, or unstable), has been well-established in tokamak experiments using
both PID and model-based control techniques [Sabbagh, 2006; Sabbagh, 2013]. As the typicalunstable, global MHD mode that appears in this situation is the RWM, the approach is typically
named RWM control. Unstable RWMs are not expected in ITER until high performance plasmas
are produced, most likely in the FPO phase of operation. However, the control algorithms, sensors,
and actuators used for DEFC and RWM control are the same, so that commissioning of the control
system for these tasks can be performed at the same time. The only difference that may occur for
RWM control is that control of faster growing unstable RWMs may require faster power supplies
than will be available in PFPO-1 for the ELM control coils.
The sensors for DEFC and RWM control are relatively low frequency (< 1 kHz) magnetic sensors,
typically partial saddle loops, that are positioned toroidally to allow real-time discrimination of n =1 magnetic field perturbations. Discrimination or higher-n numbers will be required to correct
higher-n error fields and modes as desired (or determined to be needed after the initial operation of
ITER), but correction of fields with n > 3 may not be important. Many such sensors will beavailable for PFPO-1 and later operational periods. The actuators to be used for DEFC are theexternal error field correction coils (for nearly static error field components) and the ELM control
coils (for dynamic error field components, plasma amplification of the error field, or plasma mode
activity toroidally rotating faster than the external error field correction coils can track). The key
timescale for active feedback is the time rate of change of the error field for DEFC, and the time
rate of change of the plasma amplification of such fields. When due to an unstable RWM, thistimescale is determined by the inverse of the mode growth rate. These timescales will besignificantly slower than the latency of the feedback system. However, the timescales will bridge
across the capability of the slower external coils, to the faster ELM coils.
As part of the previously mentioned initial rotation profile control development it will also be
possible to begin to develop dynamic error field correction (DEFC) [Garofalo, 2002; Sabbagh,
2010; Paz-Soldan, 2014] using the ELM control coils with a limited set of power supplies available
in this phase. This also provides an opportunity to develop initial RWM control schemes [Boozer,
1995; Garofalo, 1999; Bialek, 2001; Fitzpatrick, 2002; Okabayashi, 2005; Sabbagh, 2006;
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Reimerdes, 2007] with the limited ELM control coil power supply set, even though such control
will not be required until high performance operation is achieved.
With at least two pellet injectors and a limited set of ELM control coil power supplies, it will bepossible to begin to develop ELM control techniques in PFPO-1. While ELM control is notexpected to be required in the PFPO-1 phase because the expected heat loads for these lower
plasma current and lower auxiliary power conditions are not expected to be severe, it is important tobegin to develop ELM control techniques as soon as H-mode operation will be obtained to prepare
for higher performance operation. After the ECRH system, one ICRF antenna, and the pellet
injection systems are fully commissioned in PFPO-1, H-mode operation will be attempted as
described in section 2.5.4.9. If ELMy H-modes are achieved, then ELM control techniques withboth pellet pacing [Lang, 2004; Baylor, 2013] and application of 3-D fields [Evans, 2004; Suttrop,
2011] with the ELM control coils will be tested. An estimate of 6 run days is allocated to initial
ELM control experiments in PFPO-1.
Once the ECRH system is fully commissioned, initial tests of NTM control should be attemptedalready in PFPO-1 well before it will be required to avoid disruptions at higher plasma
performance. At 7.5 MA, 2.65 T with the full 30 MW of auxiliary heating power available in
PFPO-1, it may be possible to drive NTMs unstable, particularly if relatively high plasma
performance H-mode operation were achieved. Even at lower plasma performance, classical tearing
modes are likely to be driven under some conditions and their control with the ECRH system could
be attempted. Much previous work has been done to understand and develop NTM control on
existing devices and for ITER [La Haye, 2006; Maraschek, 2007; La Haye, 2009; Sauter, 2010;
Austin, 2011; Reich, 2012; van den Brand, 2012; van den Brand, 2013; Rapson, 2014] that givesome confidence that NTM control will be effective on ITER. The specific conditions in ITER,
however, will require significant run-time to fully develop NTM control on ITER. For the initialcommissioning of NTM control, 6 run days have been allocated in PFPO-1.
Once the ECRH and ICRF H&CD systems are fully commissioned, it will be possible to also begin
to develop sawtooth control techniques in PFPO-1 before they will be required for high plasma
current and low-q operation in PFPO-2. At 7.5 MA, 2.65 T with the full 30 MW of auxiliary
heating available in PFPO-1, since q95 ~ 3, these scenarios are likely to have relatively large
sawteeth, which would provide a good plasma scenario for initial sawtooth control commissioning.
A great deal of work has been done previously to develop an understanding of sawtooth control, todemonstrate it on existing devices with both ECRH and ICRF heating, and model sawtooth control
on ITER [Muck, 2005; Graves, 2005; Chapman, 2007; Graves, 2009; Chapman, 2009-1; Goodman,
2011; Kim, 2014]. Initial sawtooth control commissioning has been allocated 2 run days in PFPO-1.
2.5.4.8.3 Supervisory control
A key feature of the ITER PCS is advanced supervisory control, which includes exception handlingwhen a plant system or plasma event occurs that requires a real-time change in control as well asshared actuator management that attempts to optimize a limited number of actuators for multiple
simultaneous or consecutive control functions. Some initial exception handling is already required
from FP to monitor operation limits and take action if prescribed thresholds are approached such as
for coil protection. Exception handling is also used to perform real-time tuning or changes ofcontrollers through comparison with expected performance based on prior modelling. Exception
handling is an essential part of the PCS as the first line of defence for investment protection. As
plasma performance increases, so do the requirements for exception handling to optimize theavailable run-time. Some aspects of exception handling are already performed on JET and ASDEX
Upgrade. The PCS design team has published references to the ITER PCS design including95

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exception handling [Humphreys, 2015; Raupp, 2017]. An estimate of 12 run days has been
allocated to commission and develop exception handling in PFPO-1.
Once the ECRH system is fully commissioned to deliver 20 MW of power to the plasma routinely,
the initial actuator management commissioning can begin. This will include automatic switching of
power from the equatorial to one of the upper launchers as well as switching power between theupper and lower steering mirrors (USM, LSM) of each upper launcher. The control isstraightforward, but the timescales for switching (~3 s) need to be included in the algorithms andthe consequent dead time needs to be taken into account in the impact on the shared actuator control
schemes such as sawtooth and NTM control. Descriptions of the PCS architecture requirements andof how shared actuator management will work have been published [Treutterer, 2017; Humphreys,
2015]. An estimate of 7 run days has been allocated for commissioning shared actuator
management in PFPO-1.
2.5.4.8.4 Advanced control commissioning deliverables for PFPO-1
For PFPO-1, the advanced control commissioning deliverables include:











H-mode operation in plasmas at 5 MA/1.8 T including handling of L-H and H-L transitions;
Initial ELM control with pellet pacing and with a subset of the ELM control power supplies;
Initial first wall and divertor heat load protection;
Initial  and stored energy control;
Initial rotation profile control;
Minority species control for ICRF heating schemes;
Initial NTM, dynamic error field and RWM, and sawtooth control as plasmas parameters
allow;Development of more advanced actuator management functions;
Initial forecasting and event handling schemes to anticipate controllability boundaries and to
take evasive action to optimize plasma operation and avoid disruptions;
Magnetic control at plasma parameters of up to ~10 MA/5.3 T, depending on thedevelopment of disruption and runaway mitigation.

2.5.4.9 Options for H-mode operation in PFPO-1
2.5.4.9.1 Summary of scenarios
Experience operating ITER with H-mode plasmas in the first pre-nuclear phase (PFPO-1) would
provide critical information to optimize the exploitation of H-mode operation more than 2 years
later in the second pre-nuclear phase (PFPO-2), when the full complement of the H&CD systems isavailable. The primary goals of the H-mode operation campaign in PFPO-1 are to: 1) Validate thepredictions of the L-H threshold power and its scaling with density and, as far as possible, withtoroidal field, 2) Provide the first assessment of the energy and particle confinement properties of(dominant electron heated) H-mode plasmas in ITER, 3) Perform initial tests of ELM control using
a reduced set of in-vessel ELM control coils in plasma conditions for which unmitigated type-I
ELMs are not expected to cause melting of the W divertor monoblocks, 4) Validate projections of
the H-mode SOL power flux width in preparation for higher power H-mode operation in later
phases, and 5) Provide reference ITER H-mode plasma conditions to evaluate the possible effects of
TBMs on H-mode performance at 1.8 T. This will be explored in PFPO-2 together with the
optimization of the mitigation of the TBM effects with the ex-vessel error field correction coils and96

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the in-vessel ELM control coils, for which the reference of H-mode conditions in PFPO-1 are
crucial.
With this motivation, the plan for H-mode operation in the PFPO-1 phase can be summarized asfollows:
Highest priority and first H-mode attempts (31 days, if hydrogen H-mode operation is
successful): H-mode operation in hydrogen-dominated plasmas at 5 MA/1.8 TOperation to assess ripple effects on H-mode and to provide conditions for H-mode
exploration if H-mode in hydrogen is unsuccessful (10 days): H-mode operation inhelium plasma at 5 MA/1.8 T

Assessment of the L-H power threshold dependence on Bt (3 days): H-mode operationin helium plasma at 7.5 MA/2.65 T

Consistent with the Staged Approach plan, the H&CD systems available in
PFPO-1 are assumed to be 20 MW of ECRH with up to 8 gyrotrons at 104 GHz or 110 GHz,
depending on the technical solution chosen, and the remainder of, at least, 16 gyrotrons at 170 GHz,and a single ICRF antenna/system capable of providing up to 10 MW of heating power to theplasma. To achieve H-mode at these power levels requires a scenario for which the L-H power
threshold is less than 30 MW. Projections of the L-H threshold powers for ITER plasmas operated
at one-third of the maximum current and field (5 MA and 1.8 T) indicate that 30 MW could be up toa factor of 2-3 above the threshold for Helium plasmas and up to a factor of 1.5 above the thresholdfor hydrogen plasmas, if all the available heating power can be coupled to these plasmas and theplasma density is set to the value needed to minimize the threshold power, as discussed in moredetail in Appendix B. At half current and field (7.5 MA and 2.65T) the 30 MW available in PFPO-1
is marginal to exceed the threshold power in helium plasma (a factor ~ 1.1-1.5), and below thethreshold for hydrogen. In addition to the additional heating it is assumed that there will be a set of
power supplies connected to the ELM control coils so that n = 3 operation with coil current up to~30 kAt will be possible.
The total operational time for H-mode plasmas in PFPO-1 is in the range of 34 – 44 days,
depending on whether the 5 MA/1.8T hydrogen plasmas provide appropriate H-mode performance
to meet the main required goals of the experiments (H-mode characterization and exploration of
ELM control) or helium plasmas are required. Details of the expected deliverables from this initial
H-mode exploration phase in PFPO-1, the rationale for the prioritization of these first H-mode
attempts are given below. The associated R&D that could be undertaken during ITER construction
to address the outstanding issues for H-mode operation is described in section 5.1 while thetechnical issues and detailed physics risks associated with 5 MA/1.8 T operation in ITER aredescribed in Appendix F.
Note that a transition from L- to H-mode will change the plasma inductance and . These changes
will require adjustments to poloidal field currents in gap control to preserve the plasma boundary
shape and strike point positions. The response of the control system needs to be commissioned for
L-H and H-L transitions in flat-top or during current ramp-up and ramp-down phases. Limitations
on CS and PF coil currents for shape control in H-mode need to be commissioned and PCS needs tobe further optimized to cope with a sudden loss of heating to the plasma. Standard plasma rampdown scenarios following H-mode need to be developed and plasma termination scenarios for fault
conditions during H-modes need to be commissioned. The response of the plasma control system to
H-modes and ELMs, in particular VS, needs to be optimized and the limitations documented at as97

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low a plasma current as possible, to ensure that disruptions due to loss of vertical stability of theplasma do not cause melting of PFCs.
2.5.4.9.2 Motivation for low field (1.8 T) operation in PFPO-1
There are several positive aspects of operating the first H-modes in ITER at low toroidal field:

Lower power threshold for H-mode access: as mentioned in section 2.5.4.9.1 anddiscussed in more detail below and in Appendices Appendix B and Appendix F, operation
at one-third field should allow H-mode operation early in the experimental program, evenfor the relatively low 20 MW of ECRH and 10 MW of ICRF assumed to be available in
PFPO-1, which is not possible at higher fields in this phase. This is particularly true forhelium plasmas, where PLH from the reference scaling is in the range of 10-15 MW.

Investment protection during commissioning: For events that can lead to melting ofplasma-facing components, such as ELMs, disruptions, and runaway electrons, thecommissioning of control and mitigation systems should start in plasma operating regimes
where unmitigated events pose a low risk for melting. This is achieved by keeping theplasma current, toroidal field, and stored energy in the system as low as possible, whilestill matching key plasma parameters, such as q95, to those at high performance operation
and approaching as much as possible others that are known to affect ELM behaviour (e.g.
plasma collisionality). Evaluation of the expected ELM energy densities for these plasma
current levels in ITER [Polevoi, 2016] based on the EPED pedestal pressure predictions
and empirical scalings from experiment (mostly based on ASDEX-Upgrade and JET [Eich,
2017]) show that uncontrolled ELMs at 5 MA/1.8 T in ITER will not cause melting of the
top surface of the divertor W monoblocks in the high heat flux areas of the vertical targets,even for the most conservative assumptions for the pedestal plasma pressure (samepressure for H and He H-modes as for DT plasmas) and divertor ELM energy density(largest value within experimental scatter and considering largest in/out ELM deposition
asymmetries). There will also be some margin against overheating of divertor monoblock
toroidal gap edges, which has recently been identified as more problematic than previously
thought, even with the toroidal bevelling of the monoblock front surfaces, designed to
avoid heavy stationary loading on misaligned poloidal gap edges [Gunn, 2017]. This isdiscussed further in Appendix C. In addition, 5 MA H-mode plasmas are inside the zone ofacceptable operation with respect the avoidance of melting during unmitigated disruptions,as described in section 2.3.

Operation at q95=3: The ratio of plasma current to toroidal magnetic field near the edge ofthe plasma core (~q95) is a key operating parameter for ELM control via 3-D fields, since
the currents in the control coils need to be carefully tuned for a particular value of q 95.
q95 = 3 is the reference value of the baseline Q=10 inductive scenario H-mode at
15 MA/5.3 T and this can be matched by operating at 5 MA/1.8 T. The level of 5 MA is
comfortably above the 3.5 MA lower limit for good plasma position control in diverted
operation in ITER, as discussed in section 2.5.4.2.



H-mode operation at 1.8 T in the PFPO-2 campaign: if H-mode experiments at 1.8 T in
the PFPO-1 campaign proves successful, this would open the possibility of robusthydrogen H-modes in the PFPO-2 campaign (if this is marginal in PFPO-1 for H) as the
ICRF heating will be increased to 20 MW and thus allow the possibility of comparinghydrogen and helium H-modes for the same Ip/Bt in ITER. This would also allow theevaluation of the possible effects of TBMs on H-mode performance by comparing similar
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H-mode plasmas in PFPO-1 (no TBMs) and in PFPO-2 (with TBMs) at 1.8 T. This is an
important ingredient to guide the optimization of the mitigation of the possible effects of
TBM effects on H-mode plasmas with the ex-vessel error field correction coils and the invessel ELM control coils in ITER.
2.5.4.9.3 Details of the experimental plan for H-mode operation in PFPO-1
Assuming that the evaluation of all aspects of 1.8 T operation described in Appendix F issatisfactorily resolved, the program foreseen for PFPO-1 would amount to 44 days of operation.
This will be preceded by the completion of commissioning of the available H&CD systems inhydrogen L-mode plasmas at 3.5-5 MA/1.8 T (at least 13.4 MW of ECRH at 170 GHz, up to 6.7
MW of ECRH at 104-110 GHz and 10 MW of ICRF with heating pulse lengths of up to ~ 30 s).
Details of each of the experimental plans for each of the H-mode scenario considered are given
below.
2.5.4.9.3.1 First H-mode studies: initial H-mode operation in hydrogen plasma at 5 MA/1.8 T(17 days)
H-mode plasmas will be produced by applying the maximum available power (20 MW of ECRH
and 10 MW of ICRF) to 5 MA/1.8 T plasmas – this will already have been commissioned to pulse
lengths of ~ 30 s in L-mode plasmas. The first magnetic configuration to be explored will have thestandard ITER shape and a (separatrix) TF ripple of ~ -1.3%. Initial focus of these experiments willbe the achievement of an L-H transition at ne ~ 0.4nGW, the assessment of the issues related to thecoupling of ICRF in these conditions and an exploration of the density dependence of the H-mode
threshold to determine the minimum density for H-mode access. An assessment of the required
antenna-plasma distance to achieve the required coupling, including the changes of the edge density
profile following the H-mode transitions, which may require optimization of gas fuelling, and of theresulting first wall deposited power by (ICRF-accelerated) fast particles trapped in the ripple wells
will be performed for several values of target L-mode density. These experiments will involve RF
power ramps to maximum power on timescales of ~20 s followed by a constant power phase of~10 s to determine if type-I ELMy H-mode conditions can be sustained. These experiments will
provide the first H-mode plasmas in ITER, as well as first measurements of the evolution of plasma
parameters, following the L-H transition (important to define the details of H-mode experiments in
PFPO-2).
Assuming that appropriate ICRF coupling can be kept with a significant distance between theantenna and the plasma, the plasma-wall gap will be increased in several steps until a ripple level of
0.55% at the separatrix is achieved, with the corresponding changes in plasma shape to maintainq95 ~ 3 (this requires an inwards displacement of 0.4 m with respect to the reference shape, as
discussed in Appendix F). This will allow an evaluation of the optimum distance for hydrogen
H-mode access in this phase, which will be a compromise between ICRF coupling and power fluxesof ripple-trapped particles to the first wall. From the perspective of later ELM control studies, it ispreferable to maintain the plasma as close to the first wall and ELM control coils as possible, as theintensity of the applied field decreases strongly with the plasma-wall distance. The time required tocomplete this initial H-mode access assessment in hydrogen plasmas is estimated as 10 days.
It is expected that by this time some evidence of H-mode access in hydrogen plasmas will beobtained and the required power for this access versus antenna-wall distance (or separatrix TF
ripple value) and plasma density will have been characterized. If H-mode access in hydrogen is notachieved at this stage, the rest of the research on H-mode and ELM control in PFPO-1 (section99ITR-18-003

2.5.4.9.3.2) will be performed with helium plasmas, which will require the research in section
2.5.4.9.3.3 to be performed at this stage.
Even if H-mode access in hydrogen plasmas is achieved, stationary type-I ELMy H-modes might
not be sustained due to the low margin of power over threshold in hydrogen plasmas and/or to thespecific ICRF heating issues at 1.8 T. If so, helium will be introduced by the gas introduction
system (targeting small concentrations ~10%) to see if this can provide additional margin for
stationary type-I ELMy H-mode operation via the reduction of the H-mode threshold, as observed
in JET [Hillesheim, 2016]. Besides helium concentration scans, this will require a series of density
scans and antenna-wall scans analogous to those performed for hydrogen. The total time required tocomplete this assessment of mixed He-H H-mode plasmas is estimated to be 3 days.
Assuming that either pure hydrogen plasmas or mixed H-He plasmas provide stationary type-I
ELMy H-mode plasmas, the rest of the studies on type-I ELMy H-mode characterization andcontrol will continue with hydrogen-dominated plasmas. If this is not successful the research will
shift to helium H-modes and the studies under section 2.5.4.9.3.2 will be performed in helium,
following the studies in section 2.5.4.9.3.3 that would start at this stage.
This phase will conclude with the assessment of stationary type-I ELMy H-mode characteristics of
1.8 T H-modes with RF heating pulses of up to ~30 s, including a range of currents from 3.5 to
5.0 MA, approximately covering the range, q95 = 3.0-4.3, which is relevant for ITER high-Q
reference scenarios. It is expected that some properties of these H-modes will have been assessed
during the initial experiments to determine H-mode access, as described above. These dedicatedexperiments will prioritize power and density scans (ideally over ne ~ 0.2-0.9nGW) to determine theoperational range of type-I ELMy H-modes, their confinement properties, ELM behaviour andassociated power/particle fluxes to PFCs. An initial assessment of the effects of plasma fuelling
schemes on type-I ELMy H-modes will also be performed by comparing gas fuelled hydrogen Hmodes with pellet fuelled ones (two pellet injectors will be available in PFPO-1) at several plasma
density levels. This is expected to require at most 4 days of operation during which verification of
the functionality of the systems needed for gas balance will take place, in preparation for dedicatedgas balance experiments in PFPO-2. Due to the low margin of the available heating power above
the L-H transition, the ranges of these scans are expected to be limited. However, it is important toperform this stationary type-I ELMy H-mode characterization at this stage because these 1.8 T Hmodes will be the reference plasmas on which the effects of TBMs will be evaluated and mitigatedduring PFPO-2. The experimental measurements in these scans will provide the first indication of
the scaling with plasma current of the divertor power fluxes in type-I ELMy H-modes between
ELMs and during ELMs and of the pedestal plasma parameters and overall energy confinement of
H-modes in ITER, which is important for the detailed planning of H-mode experiments in PFPO-2.
As discussed in detail in Appendix C, for this level of additional heating (Pinput ≤ 30 MW) no active
measures to mitigate the divertor power loads are expected to be required for 5 MA/1.8 T plasmas.
However, some degree of control of the divertor plasma temperature may be required not to haveexcessive W sputtering; this could limit the low range of plasma densities to be explored and/or
may require the use of impurity seeding in these conditions. The precise value of the divertor
plasma temperature with acceptable W net influx will be determined by the effectiveness of W
prompt redeposition in these ITER H-mode and thus will be an issue to assess in these experiments.
2.5.4.9.3.2 First ELM control experiments 5 MA/1.8 T (14 days)
Assuming that experiments in section 2.5.4.9.3.1 deliver stationary type-I ELMy H-modes inhydrogen dominated plasmas, the next step will be to investigate ELM control in these conditions.
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Otherwise, the experiments in helium plasmas under section 2.5.4.9.3.3 will take place first andthen the ELM control experiments in this section will follow but in helium plasmas.
ELM control experiments will be performed in plasma conditions which are optimum for theachievement of stationary type-I ELMy H-modes at 5 MA/1.8 T (e.g. optimum plasma density andseparatrix-wall distance). The focus will be on the application of resonant magnetic perturbations(RMPs) with the ELM control coils, but also on the assessment of the potential of back-up ELM
triggering techniques, i.e. vertical plasma oscillations and pellet injection. This study is important at
this stage because these techniques are expected to be applied when expanding the operational
range of type-I ELMy H-modes to higher currents and fields to avoid melting of the W divertor (seesection 2.5.5.7). In fact, from preliminary estimates (to be refined in the near future), therequirements for the back-up ELM triggering techniques are not expected to be very demanding for
5 MA/1.8 T H-modes. For the vertical plasma oscillations it is estimated that the current required ineach of the in-vessel vertical stability coils to provide a 20 Hz vertical oscillation of the plasma
position by ±4 cm, which is expected to be sufficient to trigger ELMs in ITER [Gribov, 2015], is±27 kA, i.e. less than 50% of their design limit of 60 kA. Similarly, the pellet size required for ELM
triggering at this level of plasma current is expected to be 1/3 of that required to trigger ELMs for
15 MA/5.3 T Q =10 plasmas (2.01021 D atoms/pellet [Futatani, 2014]) and, thus, about an order ofmagnitude lower than the maximum pellet size that can be provided by the pellet injectors (6.21021
D atoms/pellet). The two pellet injectors available in PFPO-1 provide a combined pellet injection
frequency of up to 32 Hz, allowing a controlled increase of the ELM frequency by more than anorder of magnitude compared to that of uncontrolled ELMs [Loarte, 2014].
The assessment of the potential of the back-up ELM triggering techniques in ITER is not expected
to require a very substantial experimental time and can be possibly done in parallel with thedevelopment of ELM control by RMPs by dedicating few seconds of the stationary H-mode phaseto them. On the other hand, the development of ELM control by RMPs is foreseen to be much moretime consuming. The Staged Approach foresees a reduced set of power supplies to be connected to
the ELM control coils in PFPO-1. These will be capable of creating an n=3 magnetic field
perturbation with a similar pitch to that of the edge magnetic field for q95 = 3 as, shown in
Figure 2.5-2, by connecting up to 3 coils to each power supply. This will allow a maximum current
level of up to ~ 30 kAt in the active ELM control coils which can produce a Br/Bt perturbation for
5 MA/1.8 T plasmas similar to that for 15 MA/5.3T with the maximum baseline current level of
90 kAt. The connection of the power supplies to the ELM control coils will be configured so thatthe current level can be adjusted independently for each row of coils thus providing some capability
to vary the poloidal spectrum of the perturbation applied by choosing appropriate current levels in
the three rows of coils.
The ELM control experiments with RMPs will characterize the effects of the applied fields on theaccess and sustainment of stationary type-I ELMy H-modes at 5 MA/1.8 T, in particular: on the
power required for H-mode access and the stationary pedestal plasma parameters, ELM energy
losses and frequency, overall energy, momentum, particle and impurity confinement and power
fluxes to plasma-facing components (including divertor and first wall). This will be performed bysystematic scans of the coil current levels in the three rows of the ELM control coils and q95 scans,by variation of the plasma current, with the ultimate goal of achieving ELM suppression. For the
H-mode access phase, this will include the optimization of the timing between the applied RMP
fields and the additional heating power ramp to ensure a robust transition to H-mode while avoiding
long ELM-free periods. For the stationary type-I ELMy H-mode phase, and to the extent possible
with the reduced set of power supplies, the optimization of the RMP field applied to minimize theeffects on H-mode confinement while maintaining a given level of ELM control (including
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suppression, if possible) will be studied and the consequences for localized power fluxes to PFCs
and impurity exhaust evaluated. This will provide the first evidence of potential issues related to
localized power fluxes to PFCs, although at this plasma current level and additional heating levels it
is expected that active mitigation (e.g. time variation of the magnetic perturbation applied) [Loarte,2014] will not be required.

Figure 2.5-2: (Top) Projection of individual in-vessel ELM control coils on the poloidal and toroidal
planes. (Bottom) Connection of the ELM control coils to produce an n = 3 perturbation for ELM control in
PFPO-1 with 3 levels of current (if required) in the Upper, Equatorial and Lower row (IU, IE, IL < 30 kAt).

As far as possible, the plasma parameters of the stationary type-I ELMy H-modes will be scannedby varying input power, plasma density and fuelling scheme (gas versus pellets) and thecharacteristics of the RMP fields applied will be retuned to maintain the same level of ELM control(including suppression, if possible). This will include an evaluation of the changes to RMP required
to avoid ELM triggering by the injection of fuelling pellets, with a similar deposition profile to that
expected for 15 MA/5.3 T Q = 10 plasmas, if ELM suppression is achieved. The total time required
to perform these experiments is estimated to be 14 days. The results obtained during this phase willbe essential for the validation of models for ELM control and for the detailed planning of H-mode
experiments in PFPO-2, where ELM control may already be required to avoid divertor melting andW accumulation in higher current H-modes.

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2.5.4.9.3.3 H-mode in helium at 1.8 T and 5.0 MA (10 days)
The extent of H-mode operation in helium plasmas in this phase of PFPO-1 will depend to whichlevel the experiments in section 2.5.4.9.3.1 deliver reliable stationary type-I ELMy H-mode plasmas
in hydrogen. If this is not possible, either due to the higher H-mode threshold of hydrogen plasmas
or because of issues limiting ICRF heating at 1.8 T in hydrogen plasmas (e.g. compatibility withhigh TF ripple, etc.), the experiments to characterize type-I ELMy H-modes in 5 MA/1.8 T in
section 2.5.4.9.3.1 and the ELM control experiments in section 2.5.4.9.3.2 will be performed inhelium plasmas.
Operation in helium plasmas in PFPO-1 will require additional operational time to determine the
H-mode threshold in 5 MA/1.8 T H-modes and its scaling with density as well as to determine theeffect of TF ripple in helium H-mode access and on the sustainment of type-I ELMy H-mode
conditions. The experiments would be analogous to those in the first part of section 2.5.4.9.3.1 andare thus expected to require 10 operational days. For helium plasmas the main heating scheme to be
used is ECRH and, thus, the investigation of ripple effects on H-modes is much morestraightforward than when ICRF is required. On the other hand, if the proposed ICRF scheme basedon second harmonic hydrogen minority in helium plasmas proves viable to provide 10 MW ofheating in ITER (see Appendix F), this would allow operation in H-mode with Pinp/PL-H ~ 2.0-3.0.
This would provide a much better way to characterize the scaling of H-modes with additional
heating power level in PFPO-1 than with hydrogenic plasmas and experimental time needs to beconsidered for such studies. In addition, operating with helium plasmas in PFPO-1 gives thepossibility to evaluate the scaling of the H-mode threshold power already in PFPO-1 by comparing
results of the power required to access the H-mode with 1.8 T and 2.65 T plasmas (see section
2.5.4.9.3.4).
In view of the arguments above and the likely need to perform the type-I ELMy H-mode
characterization in helium plasmas for 7.5 MA/2.65 T (see section 2.5.5.7), for which reference
scenarios at 5 MA/1.8 T are required, 10 days of helium H-mode operation at 5 MA/1.8 T are
included in PFPO-1. It should be noted that the final number of days of helium H-mode operation in
PFPO-1 may end up being significantly longer (up to 28 days) if the initial hydrogen H-mode
development in section 2.5.4.9.3.1 is not successful. In that case the type-I ELMy H-mode
characterization (4 days in section 2.5.4.9.3.1) and the ELM control experiments (14 days in section
2.5.4.9.3.2) will be performed in helium.
2.5.4.9.3.4 Determination of the H-mode threshold in helium plasmas at ~ 2.65 T (3 days)
The H-mode threshold for helium plasmas at 7.5 MA/2.65 T is predicted to be in the range of
19-28 MW (at ne = 0.4nGW) while it is 37 MW for hydrogen plasmas. Therefore, the available
heating power (10 MW of ICRF and 20 MW of ECRH) is sufficient to attempt the determination of
the H-mode threshold scaling in helium plasmas at ~2.65 T. In this case, the ICRF scheme to be
used is hydrogen minority in helium plasmas, which can provide central heating at this field value.
The ECRH heating power, on the other hand will be on-axis for the 170 GHz operating in O-mode
providing at least 13.3 MW (off-axis for X-mode) and the remaining up to 6.7 MW of the 104110 GHz gyrotrons will be off-axis. The ripple at the plasma separatrix is 0.55% at this level oftoroidal field and, thus, ripple-related issues are not expected to be important in these conditions(see Appendix B).
These experiments will involve RF power ramps to maximum power in timescales of ~ 20 s
followed by ~ 10 s constant power phase (to determine if type-I ELMy H-mode conditions can besustained) for a range of L-mode target densities (ideally ne ~ 0.2-0.9nGW, or the highest
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achievable by gas puffing) and restricted scans of the toroidal field (~ 15%), to reduce the H-mode
power threshold if this is found to be marginal, while maintaining dominant central heating.
As these experiments will necessarily operate near the H-mode threshold, long ELM-free periods
may occur that can cause W accumulation and sudden disruption-prone H-mode terminations [deVries, 2016; Köchl, 2016]. To avoid this, the ELM back-up triggering schemes discussed in section
2.5.4.9.3.2 (vertical plasma oscillations and pellet hydrogen injection) will be applied. Both of them
are expected to be effective in triggering ELMs at this level of current and field [Gribov, 2015;
Loarte, 2014] within the design specifications of both the vertical stability coils/power supplies andof the pellet injector. This will be further verified by the demonstration of these schemes with
5 MA/1.8 T H-modes in section 2.5.4.9.3.2. It is not expected that stationary type-I ELMy
H-modes will be achieved in this phase even if the ELM back-up triggering schemes are applied. In
particular, the continuous injection of hydrogen pellets to trigger ELMs in helium plasmas will
increase the H-mode threshold towards that of hydrogen and will eventually lead to a H-Ltransition. However, the quantitative percentage of hydrogen in helium at which this will actuallyoccur in ITER remains uncertain [Ryter, 2013; Hillesheim, 2016].
2.5.4.9.4 Deliverables for the H-mode operation campaign in PFPO-1
The key deliverables expected from operation of ITER in H-mode during PFPO-1 are:

Quantification of the power required for H-mode access: The operating scenarios andplanned experiments are designed to assess the L-H transition threshold power (PLH) as
soon as possible in the experimental program. This will provide an initial assessment of Hmode access in ITER over a range of species (H and He and H-He mixtures), densities andfields (1.8 T and 2.65 T). This is essential both for the detailed planning of H-mode
operation in PFPO-2 and to determine if an upgrade of the baseline heating power level of
73 MW will be required to ensure successful DT operation. There are significant
differences between ITER operation and the data from present tokamaks, from which P LH
predictions for are made for ITER, that are known to affect the required power to accessthe H-mode. Therefore, a confirmation of the actual level of power required to access the
H-mode in ITER is essential for the rest of the experimental program and to plan power
upgrades. These would mitigate the effects of a larger than expected H-mode threshold that
could otherwise impact the start of FPO operation in ITER.



Determination of the confinement properties of H-mode plasmas and pedestal plasma
characteristics: This initial plasma operation will provide the first experimental evidence
for the performance of type-I ELMy H-mode plasmas in ITER and of the limits to the edge
plasma parameters imposed by MHD stability in ITER. The specific aspects of ITER
plasmas that will be characterized concern: stiffness of the core temperature profiles inelectron heated H-modes with no torque input (albeit at high Te/Ti), particle transport withlow core sources and gas versus pellet fuelling of ITER H-modes, edge MHD stability ofcollisionless pedestal plasmas in ITER (the expected pedestal collisionality of 5 MA/1.8 T
H-modes ne = 0.4nGW in ITER is a factor of 2.5 times lower than for 15 MA/5.3 T Q = 10
plasmas).



Characterization of ELMs (energy-particle losses and power fluxes to PFCs) anddemonstration of their control: The initial H-mode plasmas in PFPO-1 will produce thefirst type-I ELMs in ITER and their features will be investigated over a restricted range ofplasma currents, input powers and plasma densities. These experiments will be used tovalidate the predictions for the uncontrolled ELM energy losses, ELM power fluxes, and104

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impurity exhaust (including W). These are key to determine the range of plasma currents
in which uncontrolled ELMs will be acceptable in ITER and thus for the strategy to
expand the range of type-I ELMy H-modes towards higher plasma currents in PFPO-2.


Investigation of the ELM control capabilities and of the effects of RMP fields on H-mode
access and confinement: The initial set of ELM control coil power supplies will provide n= 3 magnetic field perturbations with a reduced set of poloidal spectra that allow the first
optimization studies for the application of RMP fields to H-mode plasmas in ITER to be
performed. The experiments planned in PFPO-1 will provide first evidence of the effect of
3-D fields on the power/density required for H-mode access, ELM control (with the goalof achieving ELM suppression in 5 MA/1.8 T plasmas), power/particle fluxes to thedivertor/first wall, pedestal plasma characteristics and energy, particle and momentum
transport in controlled/suppressed type-I ELM H-modes. In addition, the demonstration of
ELM triggering back-up schemes (vertical plasma oscillations and pellet pacing) will be
performed and their effect on ELMs and on the H-mode plasmas will be investigated.
ELM control validation of the reference scheme based on RMPs and of back-up schemes isvery important for the detailed planning of H-mode operation in PFPO-2 at higher fieldand current, for which uncontrolled ELMs could lead to melting of the W divertor
monoblocks.

Provide first operational experience of ICRF heating of H-mode plasmas: The applicationof ICRF for the heating of 5 MA/1.8 T H-modes will provide the first operational
experience of this heating scheme in H-mode conditions in ITER, including during the
H-mode access phase. This will allow the demonstration of the capabilities of ICRF tocouple power for the typical edge plasma conditions of H-modes in ITER, including a scanof separatrix-wall distance, both for stationary phases (possible steep density gradients in
the SOL and pedestal between ELMs) and during ELM transients, and for a range ofconditions (e.g. with uncontrolled and controlled ELMs) as well as in suppressed-ELM
H-modes. This will allow the tuning of the ICRF systems to provide efficient heating for
H-modes in ITER which will be further exploited in PFPO-2 and later phases.

Establish reference H-mode performance without effects of TBMs: The experimental
program in PFPO-1 will provide reference conditions regarding H-mode access and type-I
ELMy H-mode performance without the TBMs at 1.8 T, as they are not installed in thisphase. This information is key to evaluate the effect of the TBMs on H-mode access andperformance in PFPO-2, which, if sizeable, will then be followed by the development of
mitigation schemes utilizing the error field correction coils and the ELM control coils
within the PFPO-2 program. The potential issue with this strategy is that given theavailable power in PFPO-1, the assessment of the TBM effects will have to be done at
1.8 T in PFPO-2, as the reference plasmas will be at this field. As discussed above and in
more detail in Appendix F, 1.8 T operation in ITER has a level of ripple much higher thanat 2.65 T or 5.3 T (factors of 2.5 - 4) and the TBM themselves also create a ripple which
varies with the toroidal field. Therefore, specific experiments and analysis will be required
to separate the effect of both ripple effects when comparing plasmas at 1.8 T in PFPO-1
without the TBMs and in PFPO-2 with the TBMs. To allow this, the assessment of theeffect of TF ripple on H-mode plasmas is included in PFPO-1 (possibly in helium plasmas
that do not rely on ICRF heating, if ICRF coupling with large separatrix-wall gaps turns
out to be problematic).

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Initial validation of predictions of power fluxes in ITER H-modes: The H-mode
experiments in PFPO-1 are planned to cover a range of currents from 3.5 – 5.0 MA. This
will provide the first evidence for the magnitude of the divertor power fluxes in ITER
type-I ELMy H-modes and of the scaling of their fall-off length with current, expected toscale inversely proportional to plasma current [Goldston, 2011; Eich, 2013]. Operation
with 3-D fields applied for ELM control will provide first experimental evidence of theeffect of these fields on power fluxes to plasma-facing components (first wall and divertor)
between ELMs and in suppressed ELM regimes (if achieved) and on their dependence ofplasma parameters (chiefly density) [Schmitz, 2016]. This validation is important toconfirm and improve the physics basis for the prediction of power fluxes to PFCs in ITER
H-mode plasmas, which is required for the planning of experiments during PFPO-2 andFPO where higher levels of additional heating will be used and schemes to control thedivertor power fluxes will have to be developed.

2.5.4.9.5 Assessment of diagnostic capabilities to perform H-mode research during PFPO-1
The main objective of the PFPO-1 H-mode research program is the determination of the accessconditions (heating power, density etc.) to H-mode and the characterization of the properties of
uncontrolled and controlled/suppressed type-I ELMy H-modes. The diagnostic set considered for
PFPO-1 (see Appendix H) can provide an appropriate set of measurements to determine:Key core plasma parameters in these experiments such as ne, Te, Ti, toroidal plasma rotationand impurity levels (including W);
Power and particle fluxes to plasma-facing components as well as the influxes of recycled
species and impurity influxes.

It is likely that the time resolution of these diagnostics will not meet the requirements for
15 MA/5.3 T operation, as discussed in Appendix H, but the expected time resolution is considered
to be adequate to the objectives of this phase. Regarding edge plasma characterization the situationis less clear. For electron parameters it depends on whether or not some of the diagnostics foreseen
to be available to provide experimental measurements in PFPO-2 will already be functional towards
the end of PFPO-1, when the H-mode experiments will be performed. Of particular interest, as
discussed in Appendix H, is the early availability towards the end of PFPO-1 of the edge Thomson
scattering system, as this allows the measurements of both electron temperature and density in theedge and the determination of edge pressure gradients. In addition to this, and as a risk mitigation
measure, the ECE system could be used to determine the pedestal temperature in PFPO-1 but this is
expected to require a new radiometer to measure 2nd harmonic ECE to be added to the baseline
configuration. This, together with a peripheral density measurement from the interferometer, could
provide a measurement of the electron pressure at the pedestal, although not of its gradient at theplasma edge. Given the unavailability of the HNB and DNB systems in PFPO-1, all ion temperature
measurements will rely on the X-ray spectrometer for the core and on the X-ray crystal
spectrometer for the edge, which are expected to provide good quality measurements given the highelectron temperatures (core and edge) expected in these 5 MA/1.8T H-mode plasmas (see Appendix
F).

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2.5.4.10 Edge physics and PWI studies in PFPO-1
2.5.4.10.1 Wall Conditioning and cleaning
2.5.4.10.1.1 Glow Discharge Cleaning
Glow discharge cleaning (GDC), is one of the primary conditioning techniques, along with bakingof the vacuum vessel and blanket/first wall at 200°C and 240°C respectively, which ITER will use
to prepare in-vessel component surfaces prior to machine start-up and thereafter for plasma
operations following maintenance procedures requiring in-vessel access. The efficiency of GDC,
i.e. its ability to efficiently desaturate wall surfaces and remove medium Z impurities (adsorbed
oxygen, water or carbon) must be assessed either in H2 or in He in PFPO-1. It should be noted thatthe impact on subsequent plasmas of He-GDC (or any He-based plasmas) operated on Be PFCs has
never been assessed in a tokamak since JET with the ITER-Like Wall (ILW) has never operated inHe.
Careful tuning of GDC discharge parameters (pressure, current) will allow the optimization ofcleaning efficiency. Qualitative and quantitative assessment of the exhaust gas will be performed
using Residual Gas Analyzers (RGA), and their proper calibration beforehand is mandatory.
Spectroscopy on selected impurities (e.g. oxygen, carbon, beryllium) could also eventually beforeseen as a monitoring tool during later GDC phases when the diagnostics become fully available.
Eventually, as for all major tokamaks, routine GDC protocols will be developed, in combination
with bake-out cycles, such that conditioning is performed until designated vacuum conditions are
obtained.
The continuous follow-up of wall conditions after the first plasma, by means of optical emission
spectroscopy and mass spectrometry, and the impact of impurity levels on plasma operation andperformance during PFPO-1 will not require dedicated machine time but will be particularly
important to define further conditioning needs. This should include routine monitoring of impurity
sources in reference pulses (see Appendix D).
2.5.4.10.1.2 Ion Cyclotron Wall Conditioning/Electron Cyclotron Wall Conditioning
Ion Cyclotron Wall Conditioning (ICWC) and Electron Cyclotron Wall Conditioning (ECWC) need
to be developed for the conditioning of the first wall surfaces between pulses, i.e. when the toroidal
magnetic field is on. As stated in section 2.2.8, ICWC is preferable to ECWC, but the latter option
should be maintained in case of unavailability of the ICRF system in PFPO-1. Among the questions
to be addressed in PFPO-1, but which are common to all operation phases, is the ability of suchmagnetized low temperature plasmas to efficiently remove intrinsic (oxygen, water or carbon) andextrinsic (nitrogen, argon) impurities thus allowing/easing plasma initiation (burn-through), torecover from disruptions, and to control density in long-pulse discharges (t  200 s) with the ITER
material mix. The control of hydrogen wall content by ICWC or ECWC discharges and thereafter inensuing plasmas may turn out to be mandatory to achieve PFPO-1 and PFPO-2 targets. It istherefore preferable to develop ICWC or ECWC scenarios for each of the main toroidal field values
that are expected to be used routinely in PFPO-1. For 1.8  Bt  3.3 T and f = 170 GHz, ECWC
plasmas will be produced at the 2nd or the 3rd harmonic of the ECR frequency, and it is unclear what
the single pass absorption will be in these discharges. Since coupling of the power in ICWC ismainly non-resonant, discharges can be produced at very low Bt. The so-called high harmonic(n~10) ICWC scenario, successfully operated in ASDEX-Upgrade and in the JET-ILW, may even
therefore be envisioned in the case of failure or unavailability of GDC in ITER.

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Schemes for optimization of the cleaning efficiency are common to both techniques, proceeding
first with variations of pressure and power. Due to the relatively high densities of ICWC/ECWC
plasmas (in comparison with GDC), a very large fraction of the desorbed flux is immediately
re-ionized and re-implanted into wall surfaces and only a small fraction is exhausted through thepumping system. The discharge duty cycle must therefore be adapted to mitigate this. Finally, thepoloidal field patterns need to be optimized in both discharges at each frequency/toroidal fieldcombination (f/Bt) in order to increase uniformity, wall coverage and efficiency. As for GDC,
qualitative and quantitative assessment of the exhaust gas will require the use of the tokamakpressure measurement RGAs. Spectroscopy and visible cameras can be used to assess the areas
wetted by the discharge, whereas interferometry and ECE will permit the evaluation of basic plasma
parameters and discharge uniformity.
2.5.4.10.2 Fuel retention management
Most of the dust/tritium retention/erosion diagnostics will become available during PFPO-2. The
focus of PFPO-1 will therefore be on commissioning and testing fuel management related systems,
as well as obtaining first estimates of fuel retention and removal efficiency in hydrogen plasmas.
These may be later refined and validated during PFPO-2 through a dedicated migration/retention
experiment, including, if possible, hydrogen discharges with trace deuterium (section 2.5.5.12.2.1).
2.5.4.10.3 Gas balance in hydrogen
The most widely used technique for the evaluation of in-vessel fuel retention is the so-called ‘gasbalance’ technique, which consists of a precise accounting of the amount of injected and exhausted
fuel species. Based on measurements from all metal devices, and in particular from the JET-ILW,
the expected retention rate will be lower than 1% of the injected fuel, so that careful calibrations of
the gas injection system and of the pumping speeds of the different sub-systems will be required.
The most accurate method is to perform PVTc (Pressure-Volume-Temperature-composition)measurements on gas tanks connected to both the gas injection system and the tokamak exhaust.
Part of the required hardware will be available for PFPO-2 to allow for He3 recycling (ICRF
minority). A small upgrade is required to enable gas balance analyses, consisting mainly of theinstallation of a dedicated gas container connected to the gas injection. Making this system
available for PFPO-1 is technically possible and there appears to be no showstopper in terms ofschedule. Resources will, however, have to be identified to purchase the required hardware andensure timely installation before PFPO-1.
No dedicated operation time will be required for gas balances in PFPO-1 and measurements will be
performed parasitically when hydrogen discharges are performed requiring regeneration of thedivertor cryopumps to recover the pumped hydrogen. For the reasons mentioned in section 2.2.1,
gas balance in He is not relevant for fuel retention studies, but monitoring of the exhaust gas
composition during He discharges will provide indications on the background H sources and theirevolution with time over a campaign. Note that gas balance analysis should, if possible, be
performed at least each time a reference discharge is performed (see Appendix D) providing
information on the evolution of the wall conditions and the retention rates. Time will be required todetermine the pumping speeds of the different systems, and perform a calibration of the gas
injection system, but this can probably be done during short-term maintenance periods.

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2.5.4.10.3.1 Ammonia formation during nitrogen-seeded discharges
Although power flux densities will be too low in PFPO-1 to require impurity seeding (see Appendix
C), the development and commissioning of divertor detachment control will make use of seeded
impurities to characterize and validate the required diagnostics and actuators and the plasma
response. First studies of ammonia formation can therefore be performed during this phase
whenever nitrogen injection is performed. Recent results from ASDEX-Upgrade and laboratory
studies indicate that ammonia formation might proceed mainly from shaded areas in the divertor
region not in direct contact with the plasma, so that the ammonia formation rate might only dependon the relative fluxes of hydrogenic and nitrogen radicals to those surfaces and not too much on theactual divertor plasma conditions. In this case, a first estimate of ammonia formation rates could beobtained already during PFPO-1. This will require analysis of the exhaust gas composition using
RGAs. Calibration of these RGAs might require the injection of known amounts of ammonia into
the empty vacuum vessel. In addition, complete regeneration of the cryopumps will be necessary torecover the ammonia trapped on the active charcoal, the amount of which can be determined bychromatography of the exhaust gas. One unknown at this time, is that it has been observed in both
JET and ASDEX-Upgrade that a certain plasma operation time is required to reach a stationary
ammonia formation rate and at present it is not clear what typical discharge durations will be
required on ITER.
2.5.4.10.3.2 Baking studies
Relatively low Be erosion and co-deposition rates are predicted during the low power discharges of
PFPO-1. To first order these rates depend on the wall particle flux, itself related to the edge plasma
density/temperature. The thickness of the hydrogen containing co-deposits formed during PFPO-1
will therefore be much lower than that expected during later high-power operations. First
indications of the hydrogenic release behaviour during a bake-out of the PFCs can, however, beobtained if a bake is executed at the end of PFPO-1 (assuming that it ends with hydrogen plasmas).
In addition to providing data on the release kinetics of retained gases as the PFC temperature isincreased, such baking will allow a first validation of the global migration (including
diffusion/trapping) models used to predict T-retention in later nuclear operations. Since baking is atime-consuming activity, possible only when the magnetic field is off, these studies are best
executed at the end of PFPO-1, before the start of the subsequent assembly phase.
2.5.4.10.3.3 Development of a ‘raised strike points’ scenario
Current predictions indicate that co-deposition of Be and T in the divertor will mainly occur on theupper part of the vertical targets and on the baffle area. Given the possible limitations in theefficiency of the divertor bake for tritium removal from thick co-deposits [De Temmerman, 2017],
and the fact that divertor baking is a time-consuming activity, which can probably not be executed
too often during a campaign, it is of interest to explore techniques capable of accessing the fueltrapped in these co-deposits. Besides inter-pulse ICWC/ECWC, one possibility is to run a plasma
discharge with strike points located further up the vertical targets or even on the baffle region. This
would allow a local heating of the co-deposits, together with isotopic exchange and, possibly, somephysical sputtering of the co-deposits if the local Te is high enough.
The possibility of running ITER plasmas with raised strike-points was examined in [Kolesnikov,
2013] in the context of the old divertor strategy with carbon in the strike point regions – the idea
was to position strike points on a W region extending down from the baffle region to gain earlyexperience operating on W targets. Since then the CFC/W divertor was eliminated in favour of the
full-W variant. The study found that for strike points raised to just below the current transition
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region from the straight to curved parts of the vertical W divertor targets, L-mode plasmas up to 14
MA would be controllable in terms of vertical stability, well above the maximum plasma currents
envisaged to be attempted in PFPO-1. No analysis of the divertor pumping efficiency in these
configurations has yet been made. For detritiation studies, strike point locations on the baffle area
may even be required (providing proper attention is paid to monoblock edge overheating given thelower tolerances on block to block alignment in the baffle area). Dedicated time will be required todevelop such equilibria and to assess the maximum allowable input power etc.
2.5.4.10.4 Material migration studies
Initial studies of material (mainly Be) migration in ITER can be performed during PFPO-1. Visible
spectroscopy in the main chamber and the divertor will be available during this phase. Monitoring
the evolution of the main chamber and impurity sources requires that a standard (or reference)monitoring discharge is executed regularly during the campaign so as to provide reproducible andidentical plasma conditions. Such a reference pulse would also be useful other purposes such as
diagnostic calibration, verification of the thermal response of the plasma-facing components,
verification of the wall conditions, etc. (see Appendix D). The pulse will consist of several phasessuch as limiter, L-mode, H-mode, with scans of the separatrix location and outer wall gap distance,
allowing the study of Be transport in different plasma configurations. The discharge will evolvewith the machine capabilities, and during PFPO-1 will probably consist only of the limiter andL-mode phases. Absolute calibration of the spectroscopic diagnostics will be required, together witha regular monitoring of the transmission of the optical path, to be able to convert the measured
signal into photon fluxes which can then be compared to the signals calculated by synthetic
diagnostics of edge plasma codes.
In addition, a set of first wall samples will be installed during the assembly/maintenance phase prior
to each operation phase. They will be installed in slots between First Wall Panels (FWP) at different
toroidal and poloidal locations and will be shadowed from direct plasma impact. The aim is toevaluate the first wall erosion caused by fast charge-exchange (CX) neutrals. The sample surface
will be made of Be with marker layers to provide high resolution erosion measurements. After
PFPO-1, some (or all) of the samples will be retrieved and analyzed in a dedicated laboratories. The
data from these samples will provide useful information to validate the edge plasma/migration
models used to predict material migration.
2.5.4.10.5 Limiter heat load characterization
As in many tokamaks, plasma current ramp-up in ITER will begin in limiter configuration on the
Be first wall. This will be the case for all plasmas, in all operational phases, with the only difference
being the range of Bt which will be explored in each phase. The FWP are designed such that startup can be on either the outboard or inboard equatorial regions, with enhanced heat flux (EHF)
panels installed there to handle perpendicular heat loads up to ~4.5 MWm-2 [Mitteau, 2011]. For anumber of reasons [Kocan, 2015], inboard start-up is preferred and the majority of start-up
scenarios developed for ITER favour this route. So far, all ramp-down scenarios bring the plasma
down into a limiter configuration on EHF outboard panels, first at the equatorial midplane thenlater, in the very final stages, mostly on panels in the lower part of the poloidal cross-section at low
Ip.
In recent years, a great deal of work has been invested in the R&D community, led through the
ITPA Divertor and SOL Topical Group, to properly characterize start-up limiter heat loads. Thiswas stimulated by the first measurements on the JET ILW Be main chamber protection limiters[Nunes, 2012; Arnoux, 2013] which found strong evidence for a very narrow feature in the SOL
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parallel heat flux at the last closed flux surface (LCFS) of inboard limiter plasmas, followed by thenormal ‘broader SOL’ (of width q,main), which had previously been assumed for the design of the
ITER FWPs to be the only structure of limiter SOLs. In fact, multi-device studies [Kocan, 2015]
have revealed that the width, q,near, of this narrow feature scales with 1/Ip in the same way as forthe near-SOL in diverted H-modes (see Appendix C). It has also been shown [Goldston, 2015] to bereasonably well matched by the heuristic drift model developed for the H-mode [Goldston, 2012].
In the light of the JET findings, a very extensive multi-machine database has been constructed forthe inboard limiter q,main [Horacek, 2016] to try and confirm the original assumptions for the broad
SOL width made for the ITER wall design which were based on very sparse data. The new scalings
derived from these databases for both q,near and q,main have been used to redefine the toroidal shapeof the ITER inner wall FWPs [Kocan, 2015] since the previous major release of the ITER Research
Plan (the outer wall panel shape remains unchanged). This new shape takes into account thepresence of a double exponential structure in the SOL parallel heat flow.
A great deal more information is now also available regarding the accuracy with which the blanket
can be mechanically aligned (see Final report from the Magnetic Axis Working Group[ITER_D_SSKMMC, 2016]) and it is now even more important than before that the Research Plan
foresees adequate operational time for limiter heat load verification to carefully validate FWP heatloads, given the penalties that must be introduced to account for blanket misalignments and the factthat the real design of the FWP contour is only an approximation to the ideal shape defined byphysics [Mitteau, 2015].
Dedicated limiter heat load studies will be required in PFPO-1 at all major toroidal field values tobe used for scenario design and DMS commissioning. Part of these experiments can be conductedduring the commissioning phases that must be performed to establish routine limiter operation, as
detailed in section 2.5.4.2. The first stages will be at lower Ip (2 MA), followed by a step to near 3.5
MA, which is the approximate plasma current at which a divertor configuration can be formed
according to current scenario design. This commissioning will also include development of ECRH
heating scenarios; this is important for limiter heat load testing which requires additional power.
The early tests, probably starting at lower Bt, will permit a first crude look at the influence of FWP
misalignment on the power distribution (using the wide angle main chamber IR viewing
diagnostics) and a cross check of the magnetic axis measurements which will have been performed
following the first plasma attempts and prior to full blanket installation.
The FWP shape design has been optimized for q,main = 15 mm and 50 mm for outboard and inboard
limiter plasmas respectively and q,near = 4 mm for inboard limiter configurations (the narrow
feature is not present for plasmas limited on the outboard). The chosen values for q,main fall within
the inner wall limiter multi-machine database scatter at for Ip up to 7.5 MA, the highest plasma
current considered in the FWP design (see Heat and Nuclear Load Specifications[ITER_D_2LULDH, 2009]). Regression analysis of the database finds a best fit engineeringparameter scaling for inboard limiter plasmas of q,main  (PIN/V)-0.4-1.3 with PIN the total input
power, and V the plasma volume. For inner wall limiter configurations, a second parameter also
plays an important role in the new shape design: the effective ratio of power transported in the nearand main SOL regions of the profile. This has been fixed at Rq = 4 for the ITER inner FWP [Kocan,
2015], but is uncertain since the data from the multi-machine experiments are more scattered and ascaling cannot yet be defined. There is, however, some flexibility here since the chosen toroidal
shaping allows for a range of possible Rq before power overload, although this will be sensitively
dependent on the degree of FWP misalignment and must be carefully checked.

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Thus, according to current best scalings, there is no explicit Bt dependence in either q,main, or q,near,a relatively clear q,near  1/Ip dependence and uncertainty in the variation of Rq, though there issome evidence for a direct scaling with P IN [Horacek, 2015], so that both q,main and Rq are likely to
be input power dependent. Detailed assessment of the FWP power handling during design phases
was conducted assuming Bt = 5.3 T, so that the safety factor at the Last Closed Flux Surface (qLCFS)
will be lower at a given Ip for ramp-ups with 1.8 T and 2.65 T (e.g. for near full bore limiter
plasmas at ~1.5 at 3.0 MA, qLCFS ~ 10 (5.3 T), 5 (2.65 T) and 3.5 (1.8 T)). For a given plasma
shape, this is unfavourable for FWP loading since higher pitch angles at lower field mean that field
line penetration may occur in regions not anticipated in the original design. Detailed field line
mapping for all configurations to be tested, incorporating basic SOL heat flow models, will need tobe performed before any dedicated power loading experiments. The very first attempts at low Ip canbe used to calibrate the models and tools now being put in place to perform these assessments. It
should be noted, however, that the main chamber IR systems, which are the primary diagnostics forthis experiment, will not be sensitive to very low surface temperatures and so at very low input
power, limiter heat flux patterns may not be easy to measure.
Once limiter plasma and EC power commissioning are satisfactorily completed at each given Bt,
dedicated limiter heat load characterization in PFPO-1 should include the following main elements,
conducted for both inboard and outboard limiter configurations:Elongation scan at fixed Ip and PIN;
Power scan at fixed shape (elongation) and Ip;
Ip scan at fixed elongation and PIN;
Density scan over the range possible at fixed Ip, PIN and elongation.

The experiments should be performed in hydrogen plasmas since helium sputtering will be higher
and impurity distributions different to those in hydrogenic plasmas. Moreover, if the narrow feature
width scales with ion orbit width, this will be different in H and He. Similar experiments could be
run, if required, in He plasmas to expand the limiter loading database, but this should not be apriority in the Research Plan.
A simple criterion was adopted for the FWP shape design to define the maximum power fluxes
during limiter phases: Plimiter [MW] < Ip [MA], so that at the highest specified Ip (7.5 MA), a total of
7.5 MW of heating power (ohmic plus additional heating) is assumed. This allows for a reasonablemargin for ECRH heating, though as mentioned above, detailed wetted area and power flux density
calculations will be required prior to any limiter plasma work at higher input powers. The lower
power commissioning stages will have permitted any gross FWP panel misalignments to bequantified and this will also set the limits to what can be done during the full heat load
characterization, both with respect to input power and plasma shape variation. The maximum valuesof PIN and, in particular Ip, will need to be chosen for these experiments once the firstcommissioning results are available and will be dependent on progress with disruption mitigation(see section 2.5.4.6).
Experiments on JET with Be limiters performed long before the ITER-Like Wall installation foundno runaway generation during unmitigated disruptions up to 5 MA at high toroidal magnetic field[Harris, 1990], but the limiter configuration is more favourable for runaway appearance when
mitigation is deployed [Reux, 2015]. However, thermal loads during the current quench of
unmitigated disruptions at higher currents can be prohibitive due to conversion of magnetic energy
into thermal power flux and subsequent melt damage to the first wall [Matthews, 2016]. The choiceof maximum Ip for the limiter heat load characterization will therefore primarily be determined by112

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the status of DMS testing. If feasible, maximum target values would be ~4 MA at 1.8 T and ~6 MA
at 2.65 T. Nearly circular plasmas will likely be restricted at the higher current and powers due tounacceptable field line penetration. These are in any case of lower interest, other than to test theshaping design, since nearly circular plasmas will only ever occur at very low Ip (hence low power)
in the very early ramp-up phases.
The aim of these limiter heat load validation experiments is to set the limits on what is acceptablewith regard to first wall power loading, to check if the input physics assumptions defining the wallshape are applicable at the ITER scale, to benchmark field line tracing and wall shape parameters
and to test wall protection algorithms. The full main chamber IR diagnostic (WAVS: Visible/IR)
system must be in place to conduct this validation – experience from JET [Nunes, 2012]
demonstrated how important it is to be able to see as much of the limiter surface as possible. For theactively cooled FWPs, steady state power handling is achieved in several seconds, so that theexperiments should guarantee phases of this duration for each of the chosen scenarios. The active
cooling is advantageous in the sense that several experimental points can in principle be obtained ina single discharge, depending on the maturity of plasma shape control. Once experiments have been
performed at the first chosen value of Bt, it should be relatively straightforward to repeat the scans,or a subset, at higher toroidal field and less dedicated experimental time should be required.
2.5.4.10.6 Heat loads and detachment control during divertor operations
A proper determination of heat loads to main chamber and first wall PFCs during diverted operationand demonstration of their control, particularly in the divertor, is a key element of the PFPO phases,
in preparation for nuclear operation, when robust power flux control will be mandatory for ITER toachieve its mission goals. Providing as comprehensive a database as possible in the early
operational phases will also allow years of model development to be validated, hopefully improving
confidence that the physics of divertor power dissipation and SOL transport in ITER is wellunderstood. The main elements to be addressed are (see also section 2.2.2):







Scaling of the near SOL L-mode and inter-ELM H-mode heat flux width;
Characterization of the far-SOL width for power and particles (main chamber interactions)in L-mode and inter-ELM H-mode;
Assessment of upper FWP power handling;
Understanding divertor detachment behaviour;
Effects of 3-D magnetic fields for ELM control on divertor target power distribution;
Assessment of ELM transient divertor and main wall power loading.

Many of the required measurements will be naturally obtained during commissioning anddevelopment of scenarios in the early phases and only a relatively small number of dedicatedplasmas are likely to be required, particularly in PFPO-1, when power levels will be low. As shownin Appendix C, steady state power loads at the divertor do not challenge the heat handling capacityof the divertor monoblocks for PSOL ≲ 40 MW, which cannot be achieved until PFPO-2, and eventhen the heat load is only problematic at low divertor neutral pressure (high divertor plasma
temperature), at which W ingress may in any case be prohibitive. It is also the case that for the
maximum input power available in PFPO-1, L-mode and ELMy H-mode loads to the upper FWP in
the secondary X-point point region for the baseline equilibrium at q95 = 3 are also far below (at leasta factor 3) the maximum power handling capability of the EHF FWP #8,9 located in this area, evenfor rsep = 4 cm, the lowest primary to secondary separatrix separation allowed during the burning
plasma phase [ITER_D_TF3RCZ, 2016].

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In PFPO-1, therefore, PFC power loads will be well within their design limits, so that active control
is not required beyond the usual (accurate) control of the plasma shape (wall gaps, strike point
positions, etc.) and the respect of operational limits (see e.g. section 2.5.4.6). Specific heat load
studies in this phase can therefore be pursued without concern for PFC damage, provided the DMS
has been adequately commissioned. Experiments will rely heavily on main chamber and divertor IR
cameras, including the high resolution divertor systems viewing the inner and outer targets at single
toroidal locations.
The following studies are to be performed, if possible, in this phase: L-mode divertor heat load characterization in H and He at the different toroidal fields to be
used, including:– Power and density scans;– Current scans from the minimum (~3.5 MA) to the maximum foreseen at each Bt (e.g.
5 MA at 1.8 T, 7.5 MA at 2.65 T).
In reality, there will be a limit to the lowest input power or highest density which can be
explored since the ITER divertor is highly dissipative and too low a power or too high adensity will result in strong detachment and operation too close to disruption boundaries.
The best power load measurements will be obtained with low density/high power and withthe divertor more attached. The aim is to obtain the L-mode scaling of q,near and estimatesof the divertor power flux spreading parameter. Different combinations of Bt and Ip will also
allow characterization of the divertor wetting as a function of total incidence angle.
 H-mode divertor heat load characterization at 1.8 T: a first look at the inter-ELM q,near andfast IR measurements of unmitigated ELM heat loads, hopefully providing information on
potential issues of castellation gap edge loading and first data for comparison with ELM
energy density scalings (Appendix C).
 Assessment of upper FWP heat loads and far-SOL power and particle widths using theupper chamber viewing IR and main chamber visible spectroscopy, including:– Scans in rsep up to double null in L-mode and to rsep ~4 cm in H-mode for primarycombinations of Ip, Bt;– Density scans up to close to the density limit at fixed power in hydrogen L-modes toinvestigate SOL density profile broadening and its possible link to divertor detachment.
Similar observations also of value in H-mode, but the available parameter range islikely to be restricted in PFPO-1.
Power levels will, in general, be too low in PFPO-1 to provide serious tests of divertor heat flux
control using extrinsic seeding, except at the lowest densities (and thus high divertor plasma
temperatures) when the relevance of such seeding is questionable. At more reasonable densities, theaddition of even small quantities of impurities is likely to result in strong detachment. Prior to any
detachment control tests, an important first goal will be to characterize as best as possible thedivertor plasma itself, even in simple L-modes, to provide a database for model benchmarking toensure that the basic detachment behaviour in simple situations is properly understood. The
diagnostic tools necessary to constitute this database should be mostly available in PFPO-1 (e.g.
divertor target Langmuir probes, fast in-vessel pressure gauges, divertor target IR, divertor
bolometry and spectroscopy).
Some aspects of the divertor heat flux control methods developed at this stage can neverthelesscertainly be trialled in PFPO-1, in addition to tests, for example, of divertor reattachment provokedby local modifications to divertor gas fuelling and due to the effect of asymmetric toroidal gas
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injection. In this respect, hydrogen plasmas are more appropriate since detachment behaviour in He
is very different (see section 2.2.1) and, depending on the control sensors used, He plasmas would
not be a good testbed for later operation at high power with hydrogen isotopes as fuel. If impurity
seeding is to be tried, then it should be focused on N2, which radiates more readily in the divertor
and which has the added benefit of providing an early assessment, in PFPO-1, of potential ammonia
formation rates (sections 2.2.4, 2.5.4.10.3.1 and 2.5.5.12.2.2).
Depending on the availability of power supplies for the ELM control coils, some first measurementsof perturbed divertor heat flux footprints may also be possible in PFPO-1. Experiments in hydrogen
L-modes can already be extremely informative and can be used as benchmarks for modelling ofdivertor response to 3-D fields, which is a routine methodology in current devices. Experimentswith 3-D fields in H-modes (H or He) at 5MA/1.8 T, if possible, would provide a first chance tostudy divertor heat load patterns under conditions of ELM mitigation, but would probably not allow
relevant dissipative conditions to be accessed owing to the very low input powers.
2.5.5

Operations plan for PFPO-2

2.5.5.1 PFPO-2 Objectives
Operation at the full design parameters of 15 MA/5.3 T is essential in this phase. This ensures that
all critical systems have been commissioned prior to nuclear operation. In addition, operation in
H-mode at 7.5 MA/2.65 T ensures that key physics and control issues related to H-mode operationin DT are addressed at this stage, reducing the risk for D and DT operation in FPO. On the other
hand, H-mode operation at 5 MA/1.8 T allows the identification of TBM effects on H-modes andthe development of mitigation schemes, if required.
Routine operation of the HNB system at 33 MW for at least 50 s is necessary to support the physics
program envisioned for this and later operational phases. This includes commissioning of themagnetic field reduction system and the protection systems specific to shine-through (first wall) andfast ion loss (first wall and divertor). Both of these issues are especially important for H-mode
operation because they are the most likely factors to limit the use of the HNBs in hydrogen (and to alesser degree, helium) H-mode plasmas. Due to the high shine-through power fraction in H plasmas,
operation in H H-modes will most likely be restricted to short periods (~10 s) and to a very narrowdensity range. Thus a proper evaluation and protection of the first wall against excessive shinethrough loads is essential to carry out the H-mode research program. On the other hand, both MHD
instabilities and the application of 3-D fields for ELM control can potentially increase the fast ion
losses from NBI and lead to localized power fluxes on the PFCs. Therefore, it is essential thattogether with the increase in NBI power injected into the ITER plasmas and the development of
H-mode operation, fast ion losses are measured and the corresponding power fluxes are maintained
under control.
Routine operation of the ICRF system at 20 MW for at least 50 s is necessary to support the physics
program envisioned for this and later operational phases. In addition to the tasks in PFPO-1, thisrequires commissioning operation of both antennas simultaneously and at high power at 1.8 T in H
plasmas and at 2.65T in He plasmas.
For this operational phase, the ECRH system will have already been commissioned and will beapplied to the development of scenarios. Specific commissioning activities will be required for theapplication of the system to NTM control in H-modes, if NTMs are found to be unstable in the
PFPO-2 scenarios.

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During PFPO-2, the installed diagnostic set will be very close to its final configuration for DT
operation; the additional measurement requirements for the PFPO-2 campaign and associateddiagnostic systems to be installed prior to PFPO-2 are listed in Table 2-6. These additional
requirements expand considerably the range of physics studies that will be possible and also
provide access to expanded control capabilities, for example in relation to current drive and current
profile control. Therefore, significant effort will be dedicated to the commissioning with plasma of
the newly available diagnostics, as well as to their calibration and to the resolution of possible
inconsistencies among plasma measurements. In this phase, plasma parameters are expected toapproach comparable values to those in DT plasmas and, therefore, the demonstration of thediagnostic capabilities to support the operational program, as well as the accompanying research
program, at this stage of operation is a major objective of PFPO-2.
Since this is the last operational phase before the machine will become activated, it is important tocommission all systems required for DT operation as thoroughly as possible, so that any requiredchanges can be performed during the pre-nuclear shutdown between the PFPO-2 and the FPO
phases. This includes commissioning all required control, interlock, and safety functions.
The TBMs will be introduced during this operational phase and, therefore, specific experiments will
be performed to characterize the resultant error fields, to evaluate the effects on L-mode andTable 2-6 – Overview of additional measurement requirements and additional diagnostic
systems installed for the PFPO-2 campaign*
Measurement Requirement

Installed Diagnostic Systems

Core profiles of ne and Te

•
•
•
•

Core Ti, impurity measurements (CXRS)• Charge Exchange Resonant Spec. Core

Edge profiles of ne, Te and Ti (TS, ECE,
Reflectometry, CXRS)

•
•
•
•
•

Current profile measurements• Poloidal Polarimeter (10 channels)• Motional Stark Effect (baseline)

Divertor plasma characterization• Divertor Thomson Scattering

Fuel species• Charge Exchange Resonant Spec. Core
• Neutral Particle Analyzers

PWI/ Heat Load characterization

•
•
•
•

*

Core Plasma Thomson Scattering (full)
Reflectometry Low Field Side
Poloidal Polarimeter (10 channels)
Reflectometry High Field Side
Edge Thomson Scattering (0.85 < r/a < 1)
Charge Exchange Resonant Spec. Edge
Charge Exchange Resonant Spec. Pedestal
Reflectometry Low Field Side
Reflectometry High Field Side

Dust Monitor
Erosion Monitor
Tritium Monitor (commissioning)
Calorimetry

The table illustrates the additional diagnostics systems planned for installation prior to
PFPO-2 to meet the additional measurement requirements (when combined with the FP and
PFPO-1 systems). Certain systems partially installed prior to PFPO-1, as shown in Table 2-4, will be extended orcompleted in advance of PFPO-2 - see Appendix H for a full list of diagnostic systems planned for installation prior to
PFPO-2.

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H-mode plasmas, and to develop mitigation schemes. It is essential, if sizeable effects are found,that effective mitigation schemes are developed at this stage to ensure low disruptivity and highenergy confinement of H-mode plasmas before the plasma current, toroidal field and input powers
are increased to their nominal baseline values later in PFPO-2. This will require the re-optimizationof error field correction with the external correction coils and, possibly, additional support from theELM control coils for L-mode plasmas. Similarly, to characterize the effect of error fields on Hmodes, specific experiments to compare H-mode performance at 5 MA/1.8 T in PFPO-1 (no TBMs)
with that obtained in PFPO-2 (with TBMs) are included in this phase, together with thedevelopment of mitigation schemes (minimization of error fields with external correction coils andELM control coils), if required. The magnetic field reduction systems for the NBIs will produce
additional sources of error fields, which may also require a re-optimization of the error fieldcorrection to be carried out in this phase to mitigate their effects.
The H-mode research program in PFPO-2 will provide the physics basis and control schemes on
which to base H-mode scenario operation in FPO. By the exploration of H-modes at two toroidal
fields (1.8 T and 2.65 T) and over the range q95 ≈ 3 - 5 (Ip = 3.5 - 5 MA for 1.8 T and 5.0 - 7.5 MA
for 2.65 T) in hydrogen/ helium plasmas, the following objectives will be fulfilled:







Determination of the H-mode threshold in H and He over a density, toroidal field and q95
range allowing the evaluation of the power required for H-mode access in DT operation;
Demonstration of fully integrated H-mode scenarios including power load control, ELM
control, fuelling, core W control, which are key to the development of similar scenarios in
FPO;
Optimization of ELM control, divertor power load control and heating power mix for
H-mode plasma performance;
Demonstration of shape, ELM control and power load control on entry to and exit from
H mode phases;
Development of robust current ramp-up/down scenarios, including response to off-normal
events.

On the other hand, the L-mode program in H plasmas to 15 MA/5.3 T will demonstrate thecapabilities of the ITER tokamak and its ancillary systems to operate at their nominal design
parameters, including heating, fuelling and the capability of the diagnostic and control systems toprovide robust (disruption-free) plasma operation. Expanding the current range beyond 7.5 MA will
require qualification of the DMS in a careful step-wise fashion. Commissioning of predictive
systems for disruptions with avoidance strategies, development of an algorithm to trigger the
mitigation system, and qualification of the mitigation system at each specific current step will be
performed. The proposed step-ladder approach from 7.5 MA/2.65 T to 15 MA/5.3 T follows a
q95 = 3-4 band with changes of both Ip and Bt. This implies some variation in the heat deposition
profile that can move off-axis (up to  = 0.5) for some steps, if NBI is not used (which is likely at
the lower currents due to the high shine-through density limit in H plasmas). Therefore, W
accumulation will have to be carefully monitored and controlled; specifically the interplay between
divertor conditions (W sputtering source) and core W control will need to be assessed. If theinstalled heating systems (chiefly ECRH and ICRF) are shown to be insufficient to provide W
accumulation control, some of the specific values of the Ip/Bt step-ladder will have to be adjusted
or, eventually, a constant 5.3 T route may have to be chosen for this development. This is lessdesirable because q95 changes from 6 to 3 along this route, so that MHD stability can change
significantly at each step in Ip, and it can become more complicated to ensure stability as Ip
approaches 15 MA. In addition to the operational experience, the L-mode plasmas obtained through117

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this development will provide first experimental evidence of core transport physics of plasmas withparameters not far from those at Q = 10, which are essential to validate the understanding oftransport physics at the ITER scale and to plan D and DT operation in FPO.
If experimental time allows (in the plan an operational period of 6 days is put aside for an initialattempt), it would be desirable to extend the operation of the H&CD systems to several hundred
seconds, allowing long-pulse H-mode operation to be developed. This would provide initial
experimental evidence of long time-scale behaviour of ITER H-mode plasmas in preparation for DT
operation.
In addition to this long pulse demonstration, specific issues related to long-pulse operation in DT
will already be assessed at this stage. This includes target-q profile development in the current
ramp-up and sustainment for ~10 s in the flat-top, and assessment of the current drive capabilities of
the ITER H&CD systems, particularly NBI and ECRH, in H-mode plasmas that can provide an
initial assessment of possible issues related to the application of these schemes in FPO long-pulse
operation (e.g. triggering of Alfvén Eigenmode instabilities which could reduce NB current drive)and allow the development of schemes to mitigate them, if required.
It would also be desirable to demonstrate core MHD instability control, even if this may not be
required at this stage in the development of the ITER Research Plan, to ensure that the required
schemes are developed in advance of their application in FPO and to ensure that the necessary
upgrades are launched at this stage, if the baseline capabilities are found to be insufficient. Suchstudies would focus on the control of sawteeth and tearing modes primarily by the application of
focussed ECRH/ECCD.
Finally, as discussed in section 2.5.1, the wide range of experimental conditions and plasma
parameters expected in PFPO-2, in particular the high power and long-pulse operation aspects, will
provide access to studies of key PWI and heat load management issues in preparation for thesignificantly more challenging edge and divertor parameters expected at high fusion power during
FPO.
2.5.5.2 Plasma restart in PFPO-2 (including influence of Test Blanket Modules)
The plasma restart in PFPO-2 should aim to re-establish routine plasma operation at 7.5 MA/2.65 T
as rapidly as possible. The restart will be in hydrogen and should re-establish plasma initiation at
2.65 T with further optimization when required. It should include plasma initiation at full CS
precharge of 40 kA at 5.3 T, for the preparation of operation at 15MA/5.3T later in PFPO-2.
PCS should demonstrate routine operation at 7.5 MA/2.65 T during this restart phase byrecommissioning plasma shape control, VS control, density control with gas and pellets and event
handling. In parallel, the DMS system should be recommissioned using reference pulses from
PFPO-1, and any changes to DMS hardware or control implemented during Assembly Phase III
should be commissioned.
The interlocks for the H&CD systems should be retested and the ECRH systems and ICRF systems
that were previously commissioned in PFPO-1 should re-establish heating power at >50% of
maximum power for > 5s. The restart targets for systems that were already available in PFPO-1 will
have to be defined to show readiness for commissioning of new systems. The main target for therestart in PFPO-2 is to re-establish routine operation at 7.5MA/2.65T to allow for commissioning of
the additional H&CD installed during Assembly Phase III to be performed with plasma as early as
possible, since the further development of the research program relies heavily on the availability ofa reliable H&CD capability at a significant power level.
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2.5.5.2.1 Documenting the effect of the Test Blanket Modules (TBMs) during plasma restart
The newly installed TBMs may have an effect on magnetic measurements, plasma rotation, theplasma response to error fields and overall plasma performance, particularly in H-mode. The effecton magnetic measurements can be documented without plasma operation in the machinecommissioning before plasma operation. However, the effect of the TBMs on error fields and theircontrol requirements will have to be measured by assessing the lower density limit for plasma
operation (error field locked modes) and the effect on plasma rotation. The error field correction(commissioned in PFPO-1) will almost certainly have to be re-optimized to include the effect of the
TBMs at an early stage of operation. This error field correction re-optimization should also consider
the possible additional error fields produced by the Magnetic Field Reduction system for the NBIs.
The effect the TBMs on overall plasma performance can probably be assessed most readily by
repeating H-mode pulses from PFPO-1. Although these will be H-modes at low performance (up to
30 MW input power and 1.8T magnetic field only), the effect on pedestal parameters or plasma
rotation could be uniquely documented from these experiments.
2.5.5.3 Heating & Current Drive (H&CD) commissioning (HNB, DNB, ICRF)
The commissioning of the H&CD systems to full power and pulse lengths of at least several tens ofseconds (longer as experimental time and priorities permit) is a key goal for the PFPO-2 campaign
of operation to allow access to H-mode operation and to demonstrate the operational capability of
the H&CD systems before the transition to fusion power production. For the neutral beams,
operating at 870 keV in hydrogen during the PFPO phase, extensive commissioning of the beamline
components without plasma is foreseen during Integrated Commissioning III. Commissioning with
injection into the ITER plasmas can start after pre-commissioning of the NB Magnetic Field
Reduction system (Active Compensation and Correction Coils), to decrease the residual magnetic
fields to acceptable limits (~0.1 mT in the region between the grounded grid and the neutralizer) inorder to limit the beam deflection and to ensure maximum beam transmission with minimal losses
due to direct interception. In addition, stable operation of a standard full-bore plasma with a flat-top
diverted phase having a density above the minimum shine-through density limit must be wellestablished and the critical diagnostics for protection against unacceptable shine-through heat loadon the blanket first wall panels confirmed to operate routinely as real-time monitors (i.e. density,
impurity, and IRTV viewing the relevant first wall panels). This capability will therefore need to be
integrated into PCS and the beam interlock system. The error field correction (commissioned in
PFPO-1) will then have to be re-optimized taking into account the error field produced by the NB
Magnetic Field Reduction system, as already mentioned in section 2.5.5.2.1 above.
Most of the ECRH commissioning will have been undertaken prior first plasma and during PFPO-1.
All beamlines must be reconditioned to full power for a duration of ~50-100 s in PFPO-2. The
ECRH heating interlocks will also need to be re-established: all gyrotrons and launchers should bereconditioned at 90% of previous power levels.
ICRF commissioning:
As described in section 2.5.4.5, the ITER Research Plan assumes that the ICRF acceleration
program allows one ICRF antenna to be commissioned during PFPO-1, while the second antenna
will be commissioned during PFPO-2. Additional ICRF commissioning of both antennas will be
required during PFPO-2 due to the mutual interaction of the 2 antennas and, if H-mode operation at
1.8 T is not successful in PFPO-1, due the need to adapt the antenna coupling to the different edge
and SOL conditions in H-mode, as well as to the dynamic matching of properties and power
coupling during the L-H transition. Advanced commissioning steps described in section 2.5.4.5
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must be briefly repeated during PFPO-2 to verify that the performance is the same with different
plasma conditions. The additional ICRF commissioning steps are described below.
Additional ICRF advanced commissioning (requires ~ 60 s pulse dedicated time and thus will be
addressed as the development of long pulses allows):
 Demonstration of RF coupling control via gas puffing and control of the gap between theseparatrix and the antenna, to be performed both in L-mode and H-mode4;
 Assessment of the compatibility of impurity gas puffing, used for divertor protection, with
the ICRF coupling performance, to be carried out both in L- and H-mode plasmas;
 Operation with the two antennas simultaneously in order to test the cross-talk limits, i.e. thereciprocal effect of the antennas due to their receivers/transmitters behaviour (frequencies
are required to be different by typically 100 kHz in order to avoid interference between the
two antenna control and data acquisition systems);
 Commission ICRF for wall cleaning (ICWC).
An additional risk to those listed in section 2.5.4.5 is the possible parasitic edge resonances with
He3, if this is injected for alternative heating schemes, leading to additional wall loads. Such an
eventuality would need to be prevented by adequate operating instructions.
NB commissioning:
Commissioning of the HNB and DNB systems will be carried out at half-field in H or He plasmas,
using H beams with energies of up to 870 keV and 100 keV for the HNB and DNB respectively.
The density above the minimum shine-through density limit must be well established. The initialshine-through assessment is illustrated in Figure 2.5-3 for various densities and plasma current
levels in H and He plasmas. The commissioning time is estimated to be around 50 days and does
not need to be exclusively allocated to this activity. The NB commissioning with the plasma will be
carried out following these steps:
 First pulses would be dedicated to duct conditioning with pulse lengths ~ 0.1 s, and will be
progressively extended in steps to a maximum of 1 s, with continuous monitoring of the heatloads and pressure rise in the duct region due to beam duct degassing. A total of 50 beampulses is estimated and could be combined in a long-pulse plasma discharge with a flat-top
of 100 s depending on the progress of the conditioning and the gap required between the
beam pulses to restore the conditions to acceptable limits before the next beam pulse.
 The beams ducts for each of the three beam lines, HNB-1, HNB-2 and the DNB will initiallybe individually conditioned. When the HNB-1 and DNB systems are operated in parallel,
several extra conditioning pulses may be required to condition the cross over region in theduct. This may also be the case for the shared vacuum vessel pressure suppression system(VVPSS) region between the HNB-1 and HNB-2.
 The pulse length will be progressively extended from 1 s to 2 s, 3 s, 5 s and 10 s, once theduct is conditioned.
4

The RF coupling is usually better in L-mode than in H-mode plasmas, due to smaller density gradients. Optimizing
the matching algorithm settings for the L- to H-mode transition, during which the coupling varies rapidly, will requiredue attention, since the H-mode access is bound to depend on the coupled ICRF power. The acquisition of sufficientexperimental data will allow optimization of the algorithm by better anticipation of the transition. The fine-tuning of thematching will be carried out during the commissioning. A database of matched configurations should be establishedbased on this commissioning phase.

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 A range of different tilted beam angles need to be tested in case that it is required to change
them for different scenarios, checking the associated shine-through and power density
limits, but this implies only a small number of additional pulses.
 Various beam energies need to be commissioned, since the beam energy affects the
re-ionization power on the beamline components.

Figure 2.5-3: Shine-through limits for a range of densities/ plasma currents and various HNB beamenergies and species (870 keV H beams for H/He plasmas and 1 MeV D beams for D/DT plasmas).

 The power modulation capabilities of the HNB system need to be tested during thiscommissioning period.
A similar conditioning process will be repeated for deuterium plasmas in the FPO phase, but with D
beams at energies of up to 1 MeV.
The main risk during the HNB commissioning activity arises from the shine-through losses, which
are greatest at low density, but losses to the blanket first wall can also be increased by prompt orbit
losses, enhanced at low plasma current due to the reduced confinement. Therefore, as noted above,
HNB commissioning requires real-time heat load measurements on the first wall, together with thedensity profile, impurity concentration and (if possible) fast ion loss measurements.
2.5.5.4 Diagnostics commissioning and validation
Many more core and edge diagnostics will be available in PFPO-2 (see Table 2-6 and Table H-3 inAppendix H), as well as the HNB and DNB, which will allow MSE and CXRS measurements to be
implemented. This will allow significantly improved measurements to be made of the higher
performance plasmas that will be performed in this phase and require substantial commissioning
and validation of the measurements.
A number of dedicated plasma sessions will be required not only for the commissioning of the new
diagnostics but also for the calibration of some of the previously installed set due to new plasma
conditions.

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2.5.5.5 Disruption management program in PFPO-2
The disruption program in PFPO-2 will mainly focus on adapting disruption prediction, avoidance
and forecasting capabilities, and optimizing the mitigation schemes for high current operation andfor H-mode operation.
2.5.5.5.1 Disruption Mitigation System (DMS) commissioning/optimization
The DMS will be optimized towards higher currents and higher thermal energies, including
assessment of eddy current forces during fast current quenches. The latter will aim at maximizing –within the load limits – the radiated power fraction during the thermal quench in preparation tothermal load mitigation and runaway suppression. Thermal load mitigation has to be confirmed for
VDEs and MDs and scalings on thermal energy, impurity species and quantity need to beestablished. Radiation peaking has to be reduced to stay below melting thresholds when
extrapolating towards full performance pulses. This will require amongst others to optimize thesequence of all injectors.
RE avoidance has highest priority when approaching high current operation and the injection
schemes have to ensure sufficient margin to RE seed formation during the thermal quench.
However, testing this margin bears significant risks at high currents. Therefore, the constraints on
the injection quantities and sequences will have to rely to a great extent on modelling that was
already validated at lower current levels. Nevertheless, uncertainties will be high due to the
exponential growth in avalanche multiplication when increasing the plasma current.
DMS reliability needs will increase with current and energy. This will require more focus on robust
injection schemes that also consider failure of an injector during the injection sequence and possible
recovery scenarios. The PCS algorithms also will have to be updated to react on any loss of DMS
injectors during the plasma pulse, either by changing the pulse schedule or by choosing back-up
injectors.
The following Milestones have to be achieved in PFPO-2:
2.5.5.5.1.1 Confirmation of EM load mitigation at 15 MA
Prerequisite for operating at 15 MA.
Experimental strategy:Monitor mitigation performance during ‘natural’ mitigated disruptions, possibly hold points
if data is insufficient

Possibly, perform deliberate mitigated disruptions at currents above 7.5 MA if database is
not sufficient to allow scaling to the next current level

2.5.5.5.1.2 Confirmation of thermal quench heat load mitigation in H-mode
Prerequisite to run routinely at H-mode thermal energy levels.
Experimental strategy:Produce target plasmas with varying thermal energy in H-modeConfirm or adapt required quantities and injection scheme to achieve high radiation
fractions during the thermal quenchValidate compatibility with RE avoidance scheme and EM load mitigation scheme122

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The estimated number of pulses/days required for DMS commissioning/optimization in
PFPO-2 is 260/20.
2.5.5.5.2 Disruption Load Validation
Most of the work on load validation is expected to be carried out in PFPO-1. However, certain heatload validation may require operating in H-mode or at higher currents. Further unmitigateddisruptions may have to be performed if the accuracy of the electromagnetic load scalings derived
from up to 7.5 MA operation is not sufficient. Establishing reliable scaling for thermal loads
occurring during the thermal quench at higher stored thermal energies in L-mode during VDEs and
MDs may also require operation at higher currents. The load scalings are important for defining
target values for disruption rate and mitigation rate.
The estimated number of pulses/days required for Disruption Load Validation in PFPO-2 is
10/1.
2.5.5.5.3 Disruption Mitigation System (DMS) trigger generation
Start of H-mode operation will require adapting disruption prediction and possibly also detection
schemes. Also the algorithms within the PCS determining the appropriate injection sequence ineach phase during a pulse will have to be adapted to the new plasma scenarios. Any tests ofprediction or detection schemes that include deliberate disruptions have to be performed using thedisruption mitigation system. Extended diagnostic capabilities will have to be taken into accountand exploited for the prediction algorithms. Note that the required mitigation rates will also dependon the tendency of unmitigated disruptions to form runaway electrons. For higher thermal energies,
the impurity influx especially from the divertor can lead to faster thermal and current quenches,facilitating RE formation.
The following Milestones have to be achieved in PFPO-2:
2.5.5.5.3.1 Thermal quench prediction ready for FPO
Prerequisite for FPO. The required reliability depends on the expected loads and might vary for
different thermal energies (c.f. sections 2.3 and 2.5.4.6.2).
Experimental strategy:Monitor the performance of disruption prediction algorithms;

If required expand threshold and physics model-based algorithms for predicting the thermal
quench during H-mode operation, implement statistical methods in parallel;

Perform dedicated mitigated disruptions if required to optimize the disruption predictor;

Implement and test decision schemes in the PCS for the injection sequence.

2.5.5.5.3.2 Injection sequences ready for FPO
Prerequisite for FPO.
Experimental strategy:Optimize injection sequences for pre-TQ injection;Verify RE avoidance, TQ and CQ mitigation efficiency.

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The estimated number of pulses/days required to establish the DMS trigger function in
PFPO-2 is 75/6.
2.5.5.6 Advanced control commissioning in PFPO-2
With the additional systems that will be available in PFPO-2, many advanced control schemes willneed to be commissioned as soon as the associated actuator and diagnostic systems required fortheir implementation are commissioned. Since this is the last operational phase before the tokamak
structures will become activated in the FPO phase, it is important to commission as many advanced
control schemes as possible so that any changes in the hardware and software can be carried out
prior to the FPO phase. Note that all disruption related control in PFPO-2 is described in section
2.5.5.5. The run-time estimates given here are for commissioning of advanced control schemes, but
further development of the physics R&D needed to develop robust control is likely to requiresignificantly more run-time as part of the physics program. The main new actuators in this phaseinclude additional pellet injectors both for fuelling and pellet pacing, the second ICRF antenna to
inject 20 MW of ICRF power, and two HNB systems to provide 33 MW of NB power, as well as
the remainder of the ELM coil power supplies to provide full independent control of the ELM coils.
The installation of the first six ferromagnetic TBMs will also require recommissioning of error fieldcorrection. The development of more sophisticated actuator sharing, event handling, integrated
control, and forecasting techniques will also be required in this phase.
2.5.5.6.1 Magnetic control
The development of plasma current, position, shape, and vertical stability control will have to beextended during the PFPO-2 phase up to 15 MA and 5.3 T. While this will build on the magnetic
control that was developed during PFPO-1 at currents of perhaps 10 MA, because of the increased
risk of damage to in-vessel components at higher plasma current, great care will have to be taken toraise the plasma current gradually to 15 MA while ensuring the effectiveness of disruption andrunaway mitigation techniques. Increased attention will also need to be paid to plasma shape andposition control at higher plasma current.
2.5.5.6.2 Plasma kinetic control
While first wall protection will have already been developed in PFPO-1 at auxiliary heating powers
of up to 30 MW and to plasma currents of at least 7.5 MA, in the PFPO-2 phase it will need to be
developed further for total auxiliary heating powers of up to 73 MW and plasma currents of up to
15 MA. As part of HNB commissioning in this phase, first wall protection against NB shinethrough will need to be developed. This will be carried out by gradually increasing NB energy (andpower) and pulse length. The plasma current will be gradually raised to 15 MA and the first wall
protection must include this additional ohmic heat flux at low q95. An additional 8 days of first wall
protection commissioning is estimated in PFPO-2 as the H&CD systems are commissioned to fullpower and the plasma current is raised to 15 MA.
Initial divertor protection commissioning was carried out in PFPO-1, but the relatively low
auxiliary heating power in that phase is not expected to challenge the divertor. Only above 40 MW
of auxiliary heating power is the divertor heat flux protection expected to be required. As the
H&CD systems are commissioned to full power in PFPO-2, divertor protection will become
essential. See sections 2.5.5.12.3, 2.5.5.11.4, and Appendix C for more details. An additional 7 daysof divertor protection commissioning are estimated for PFPO-2.

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The sensors used for plasma stored energy and  control in PFPO-2 are expected to remain thesame as those commissioned in the PFPO-1 period. However, the actuators will change
substantially since HNB and additional ICRF will be added providing a total of 73 MW of auxiliary
heating power. Algorithms that divide the mix of the H&CD systems used (e.g. ECRH, ICRF, and
HNB) will need to be formulated, implemented, and tested. Any model-based algorithms
implemented in the PFPO-1 period will need to be upgraded to include all three H&CD system
actuators and commission H&CD system actuator sharing. An estimate of 4 run days has been
allocated to and stored energy control in PFPO-2.
Initial temperature and pressure profile control will be commissioned in the PFPO-2 phase since theprofile diagnostics and full H&CD systems will then be available. A real-time evaluation of theplasma pressure profile can more sensitively determine instability thresholds for MHD modes suchas ELMs, NTMs, kinks, or RWMs when compared to using plasma  (the volume integral of theplasma pressure profile) alone. The sensors used in such a control system will be direct real-timemeasurements of the plasma temperature profile or the plasma pressure profile based on
measurements of plasma temperature, density, and the profile of the effective plasma charge state,
Zeff. The real-time evaluation of the pressure profile peaking factor p(0)/<p>, together with a modelof the direct measurement, might be sufficient initially. Actuators for plasma temperature andpressure profile control would include the heating system actuators mentioned earlier, but now the
localization of the heating would be changed to provide the profile alteration. The control
algorithms are expected to range from simple limits to temperature and pressure gradients, suitably
normalized to yield improved estimates of instability limits, to more comprehensive stability
physics models that are still computable in real-time, or precomputed ‘stability maps’ that can beevaluated simply, or via real-time application of neural networks in the PCS. An estimate of 2 rundays has been allocated for initial temperature and pressure profile control commissioning in
PFPO-2.
Rotation profile control will be upgraded from the system commissioned in PFPO-1, now including
new capabilities. Sensors will include real-time evaluation of the plasma rotation profile from thecharge exchange diagnostic which will be possible with the availability of the NB systems.
Inclusion of new sensors will require their implementation into the PCS and further systemcommissioning time. ITER NBI modelling does not predict that this system will provide sufficient
torque on the plasma to be used as an actuator. Instead, the plasmas operated at higher plasma
current and toroidal field in PFPO-2 are expected to have increased intrinsic rotation. The physics
understanding of intrinsic rotation based on the data taken in ITER in both PFPO-1 and PFPO-2
may produce a more accurate real-time model of this effect that could be used to upgrade thecontroller during PFPO-2 if deemed necessary. An estimate of 2 run days has been allocated for
additional rotation profile control commissioning in PFPO-2.
Initial current density profile control will be commissioned in PFPO-2 with the additional Neutral
Beam Current Drive (NBCD) available and the current density profile measurements that will also
be available with the HNB and associated diagnostics. This will allow tailoring of the current
density profile late in the current rise once NB shine-through can be avoided as well as some
control of the radial locations of the q=1.5 and q=2 surfaces for improved NTM control. An
estimate of 2 run days has been allocated for initial current density profile control commissioning in
PFPO-2.
Once all three heating systems have been commissioned, it will be possible to simulate fusion burncontrol by using one heating system as an actuator to simulate fusion power. Such experiments havebeen carried out on a number of existing tokamaks [Jones, 2001; Noterdaeme, 2003; Mayoral,
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2004; Shimomura, 2007; Hawryluk, 2014], providing a strong basis for these initial burn control
simulation experiments on ITER. An estimate of 2 run days has been allocated for initial burncontrol commissioning in PFPO-2.
2.5.5.6.3 Magnetohydrodynamics (MHD) and error field control
With a second pellet injector cask with two additional pellet injectors, the full set of ELM control
coil power supplies, and the full 73 MW heating power, as well as additional diagnostics to measure
ELMs, it will be possible to further commission ELM control techniques in PFPO-2. Having 4pellet injectors will allow pellet pacing with two injectors while fuelling with the other two
injectors. The full ELM control coil power supply set will allow n=3 and n=4 configurations of theELM control coils and allow rotating the ELM coil perturbations to spread out the ELM heat flux.
With the full 73 MW auxiliary heating power, it should be possible to obtain type-I ELMy H-modes
in helium in PFPO-2. An additional 6 run days have been allocated to commission ELM control in
PFPO-2.
The DEFC and RWM control system initially commissioned during PFPO-1 will provide importantpractical operational experience for upgrading and recommissioning the system for PFPO-2. The
ferromagnetic TBMs added during this phase may significantly change the intrinsic and dynamic
error fields of the device. Also, the additional heating systems will produce greater plasma stored
energy and , which will increase plasma amplification of error fields and may begin to approach
MHD stability limits. Sensor changes during this period include the acquisition and real-time
analysis of existing magnetic sensors for detection of n > 1 error fields and possible plasma
amplification of these fields. Additional power supplies for the ELM coils will extend thecapabilities of the actuators for this task, making the control system more effective. A model-based
control system should be upgraded to include the added TBM modules, NB Magnetic Field
Reduction system, sensors, and actuator capabilities. An additional 5 run days have been allocatedto commission DEFC and RWM control in PFPO-2.
With the full 73 MW additional heating available in PFPO-2, it will be possible to use the HNB and
ICRF systems to increase the plasma  while using the ECRH system for NTM control. While theexpected  values of around unity may still be too low to demonstrate full NTM control in PFPO-2,
it is likely that classical tearing modes will appear in the current rise as well as in the flattop, which
will allow additional commissioning of NTM control techniques. NTM control is not yet a routine
tool for disruption prevention/delay in present tokamaks and may require significant further
development on ITER, depending on how effective NTM control schemes developed on existing
devices prove to be on ITER. This is partially due to its complexity, but also to the oversubscription of the ECRH systems in present devices. Additionally, the temperature at the q=2
surface on existing devices is rather low compared to that expected in ITER, requiring large power
levels to drive the necessary current. The concepts for control have been successfully demonstrated,but there remains room for significant optimization; in addition, existing schemes are, to someextent, specific to the launcher geometry of each machine. An additional 5 run days have been
allocated to commission NTM control in PFPO-2, but further run-time is likely to be required aspart of the physics program.
Having the full 73 MW of additional heating and, in particular, the ability to drive fast ions withboth ICRF and NBI, it will be possible to begin to excite Alfvén Eigenmodes both in the current
rise and in the flattop. While Alfvén Eigenmode control techniques are still the subject of R&D on
existing devices, it is expected that current density and pressure profile control techniques will also
allow some control of Alfvén Eigenmodes. Some control or influence of ECRH on Reversed Shear
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Alfvén Eigenmodes (RSAEs) and Toroidal Alfvén Eigenmodes (TAEs) has been observed on
existing devices [Van Zeeland, 2009; Garcia-Munoz, 2015; Van Zeeland, 2009] and should befurther developed for future use on ITER. It is not expected that the fast ion population will be solarge in PFPO-2 that Alfvén Eigenmode control will be required in this phase, but it will be
important to begin to commission Alfvén Eigenmode control techniques before such control isrequired in the FPO phase. While no dedicated run days have been allocated to Alfvén Eigenmode
control in the present plan, it is expected that such experiments could be carried out during thecurrent density and pressure profile control experiments in PFPO-2.
2.5.5.6.4 Supervisory control
As the actuator and diagnostic systems and plasma performance increase in PFPO-2, additional
commissioning of advanced supervisory control within the PCS will be required for the first line ofdefence for investment protection as well as for managing shared actuators across multiple control
functions.
Additional exception handling commissioning will be required to deal with potential plant systemfaults and plasma events as the number of plant systems and plasma performance increase. With thefull 73 MW available from all three H&CD systems, a fault in one H&CD system could be
mitigated by using another system, under certain conditions, for some control functions. Similarly, afault in one diagnostic channel or system could be mitigated by using a redundant channel oranother diagnostic system. Anticipating plasma events such as an H-L transition to prepare for the
resulting drop in plasma  and radial shift in plasma position need to be commissioned as part of
exception handling [Raupp, 2017]. An additional 5 run days have been allocated to PCS
commissioning in PFPO-2 that will include exception handling.
With the full 73 MW additional heating available from all three H&CD systems, more sophisticatedactuator sharing techniques will need to be commissioned in PFPO-2. As the plasma performanceincreases, more instabilities are likely to appear and more simultaneous control schemes withmultiple actuators will be developed. Preparation for simultaneous control of NTMs at q=1.5 and
q=2, as well as sawtooth control and central electron heating for impurity control will need to be
demonstrated in PFPO-2. As each of these control schemes is commissioned, it will become
necessary to also develop shared actuator control with priorities adjusted in real-time for eachcontrol function.
2.5.5.6.5 Advanced control commissioning deliverables
The deliverables for advanced control commissioning during the PFPO-2 phase include:











Plasma current, position, shape, and vertical stability control to 15 MA/5.3 T;
Density control with gas and pellet injection up to 15 MA/5.3 T;
Disruption avoidance, detection, and prediction schemes to 15 MA;
HNB system operation (2 beamlines) at powers of up to 33 MW for at least several tens ofseconds;
ICRF system (2 antennas) operation at powers of up to 20 MW for at least several tens ofseconds;
First wall and divertor heat flux control to full auxiliary heating power;
All H&CD systems operating simultaneously at injected powers of up to 73 MW;
Actuator sharing, event handling, and forecasting systems;
H-mode operation with L-H and H-L transition control;
ELM control coils and pellet pacing for ELM control;127

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





Error field correction in the presence of the TBMs;
Error field correction in the presence of the NB Magnetic Field Reduction system;
Sawtooth and NTM control; and pressure profile control;
Temperature, density, impurity, rotation, and current density profile control.

2.5.5.7 H-mode Studies in PFPO-2
2.5.5.7.1 Summary of scenarios
The development of the remaining aspects of H-mode operation in the non-active phase will becompleted during PFPO-2. This will be a more extensive experimental program (a factor of ~3times longer) than in PFPO-1 because of the substantial improvement of the capabilities of ITER
for H-mode operation and improved diagnostics. These include the full complement of the baseline
heating systems and power supplies for the ELM control coils and improved diagnostics for edge
plasma parameters (e.g. reflectometers and edge Thomson scattering, if not operational towards theend of PFPO-1) and ion temperature (e.g. charge-exchange spectroscopy), as detailed in Appendix
H. To minimize the risks of initial H-mode operation in FPO, it is important that both operational
and physics issues of H-mode operation in ITER are addressed up to a plasma current level of 7.5
MA. This would allow H-mode operation in FPO to start at this level of current and proceed
upwards in DD and then DT. The main goals of the PFPO-2 phase are:
1. confirm the predictions of the H-mode power threshold and power required for sustainmentof high confinement type-I ELMy H-modes over a range of Ip, Bt, plasma densities andspecies (hydrogen and, most likely, helium);
2. establish stationary type-I ELMy H-modes over a range of Ip, Bt, plasma densities andpowers in order to: a) characterize the scaling of H-mode energy and particle confinement in
ITER, b) characterize ELM power fluxes to plasma-facing components, c) characterize
effects of ELMs on W production and impurity exhaust that may establish limitations to theuncontrolled ELM energy loss/frequency, d) to complete the initial assessment of the scalingof the SOL power flux with a wider range of toroidal fields, plasma currents, densities andinput powers and e) to extend their duration to identify possible issues related to long-pulse
operation such as long timescale recycling issues or transport changes due to the global
relaxation of the current profile;
3. demonstrate ELM control over a range of q95 > 3 (possibly up to 4.5) with 3-D magnetic
fields for both ELM transient power exhaust and impurity exhaust and to assess the potentialof the back-up ELM triggering schemes for Ip > 5 MA H-mode operation;
4. determine the effects of the TBMs on H-mode plasma and performance by comparison with
PFPO-1 experiments and, if sizeable, to mitigate their effects by the application ofcorrecting fields with the error field correction coils and the ELM control coils first at 1.8 T
and then at 2.65 T.
The total operational time for H-mode experiments in PFPO-2 is 126 days; the experimental planfor H-mode plasmas in PFPO-2 can be summarized as follows: H-mode operation to connect with PFPO-1, evaluate/mitigate TBM effects and extend
H-mode operation at 1.8 T (40 days): This will require the use of the same species as in
PFPO-1 (hydrogen or helium) and currents (mostly 5 MA) plus additional experiments to
exploit the increased additional heating, enhanced ELM control capabilities and improved
diagnostic of the plasma.
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 Expansion of H-mode operation towards 7.5 MA/2.65 T (65-80 days): This will include
an exploration of H-mode hydrogen plasma operational range at 2.65 T and up to 7.5 MA,
including the use of small proportions of helium and impurities to expand this range,
followed by the characterization of H-mode plasmas and optimization of ELM control. The
latter will be performed in helium plasmas if the hydrogen stationary H-mode development
proves unsuccessful. Alternatively, assuming that R&D in present experiments confirms that
this is a viable solution for ITER, other options for hydrogen operation at 3.0-3.3 T withICRF three ion heating (He3 + He4) maybe considered to replace helium plasmas.
 Development of long-pulse H-modes (6 days): this will require H-mode plasmas inhydrogen at 1.8 T or 2.65 T depending on the outcome of the research in PFPO-2.
Consistent with the Staged Approach plan, the H&CD systems available in PFPO-1 will be: 20 MW
ECRH (with, at most, 6.7 MW at 104-110 GHz and, at least, 13.3 GHz at 170 GHz), 20 MW ICRF,
and 33 MW of neutral beams (HNB) with hydrogen neutrals at 870 keV. To a large degree, theplasma conditions to be explored in PFPO-2 will be determined by the power that can be coupled tothe plasma in the range of fields that will be investigated in H-mode and by the H-mode thresholdof the plasmas to be explored.
For ICRF heating of hydrogen plasmas there is a good scenario with strong SPA for 1.8 T (second
harmonic majority hydrogen) but not for 2.65 T, while for helium plasmas there is a good scenario
with strong SPA at 2.65 T (first harmonic minority hydrogen) and, possibly (this will be assessed in
PFPO-1), at 1.8 T (second harmonic minority hydrogen). As mentioned above, other ICRF schemeswith strong SPA in hydrogen plasmas are possible by using the newly demonstrated three-ion[Kazakov, 2015] scheme demonstrated in JET and Alcator C-Mod [Kazakov, 2016]. This may be
viable at 3.0 or 3.3 T in ITER, although its demonstration for the specific ions to be used in ITER
remains outstanding and will imply strongly off-axis ICRF power deposition profiles, as discussed
in section 2.5.5.7.4.6. For these reasons, this possibility is considered as an option at present and, if
found viable for ITER, may be explored within PFPO-2 instead of the 2.65 T hydrogen and helium
plasmas considered in the reference operational plan below.
Similarly, the application of NB heating to low density H-modes in this phase will be limited byshine-through loads on the blanket shield module that intersects the beams at the outer first wall,
which is required to have a lifetime of at least 30,000 pulses. Two approaches can be taken to
circumvent this issue: one is to shorten the HNB pulse and the other to decrease the HNB energy,
although this implies a reduction of the HNB power with EHNB2.5 due to the perveance matching
constraint [Singh, 2017]. Detailed studies have been carried out to determine the power levels thatcan be applied in long pulses (> 100s) with acceptable shine-through loads for a range of plasma
densities and of the NB pulse duration at maximum power when operating under these densities forhydrogen and helium plasmas [Singh, 2017; Kim, 2016] and the key results are shown in
Figure 2.5-4 and Figure 2.5-5, together with the expected H-mode thresholds for 5 MA/1.8 T and
7.5 MA/2.65 T in hydrogen and helium plasmas. As shown in these figures, for both toroidal fieldvalues, significant power in addition to that from the HNB is required to access H-mode plasmas inhydrogen, while the required additional power is much lower for helium. Access to H-mode for
7.5 MA/2.65 T in hydrogen will be possible with HNB and ECRH alone, but only by operating at
relatively low densities and, thus, for NB power durations of less than 40 s.
Within the commissioning activities at the start of PFPO-2 the shine-through power deposition will
be assessed with L-mode plasmas first in hydrogen and then in helium, as described in section
2.5.5.3. If this assessment were to show that the predictions substantially overestimate the shinethrough power deposition levels, this would enlarge the hydrogen H-mode operational space at129

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1.8 T and 2.65 T in ITER. On the other hand, if the predictions would underestimate the real levelof shine-through loads this would eliminate the possibility of H-mode operation at 2.65 T in ITER
for hydrogen plasmas. The experimental plan for PFPO-2 described below is developed according
to the shine-through load levels as predicted and to the H-mode threshold levels described inAppendix B. The plan would, therefore, need to be adjusted depending on the experimental results
concerning the H-mode access powers and the shine-through power loads determined byexperiments in ITER itself.
The total operational time for H-mode experiments in PFPO-2 is 126 days. Details of thedeliverables from these plasmas, the rationale for the prioritization of the H-mode mode researchand the risks associated with each phase are discussed below. The associated R&D that could be
performed during ITER construction to refine the H-mode plans during PFPO-2 is discussed in
section 5.1.
Hydrogen plasma
75

Helium plasma

unlimited duration

PNBI

75
unlimited duration

PNBI

50

PL-H @ 2.65T (7.5 MA)

PL-H @ 1.8T (5 MA)Power (MW)Power (MW)PL-H @ 2.65T (7.5 MA)

25

0
0.0

2.0x1019

4.0x1019

6.0x101950

PL-H @ 1.8T (5 MA)

25

0
0.0

2.0x1019

4.0x1019

6.0x1019-3<ne> (m )-3<ne> (m )Figure 2.5-4: Maximum heating power which can be applied with unlimited duration versus plasma densitywith acceptable are shine-through loads for plasmas at 5 MA/1.8 T and 7.5 MA/2.65 T in (left) hydrogenand (right) helium. The red and blue lines (or shaded areas) show the expected H-mode threshold power forthese plasma conditions and the density range of ne = 0.4 - 1.0nGW, where nGW (5 MA) = 4.01019 m-3 andnGW (7.5 MA) = 6.01019 m-3.

130

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Figure 2.5-5: Duration of the NBI heating pulse at maximum power (33 MW at 870 keV) versus plasma
density with acceptable shine-through loads for (left) hydrogen and (right) helium plasmas at 7.5
MA/2.65 T. The vertical red dashed line corresponds to the density for which NBI at full power can beapplied without any limitation in heating pulse length and the vertical blue line to ne = 0.4nGW; fordensities lower than these the scaling providing the H-mode threshold power in ITER is not expected to bevalid.

2.5.5.7.2 Motivation for H-mode scenarios in PFPO-2
H-mode operation during PFPO-2 aims at characterizing this scenario and developing the control
schemes that will be required up to currents of ~7.5 MA, which are foreseen for the start of
H-mode operation in FPO. During PFPO-2 all systems required to perform this research will beavailable in ITER in the final baseline configuration including heating and current drive, pellet
fuelling, diagnostics (except most of the fast particle and some neutron diagnostics) and ELM
control coils. In elaborating the PFPO-2 plan, it is assumed that in PFPO-1: an initial assessment of
the additional heating power level to access H-mode in ITER will be performed, H-mode operation
at 1.8 T will be demonstrated in a reduced operational range, and initial studies of ELM control
with RMP fields will be performed. The availability of all the required systems in PFPO-2 willallow a wider exploration of H-mode plasmas in hydrogen and helium to be performed in
conditions closer to those in FPO, to perform a first systematic study of H-mode performance in
ITER plasmas with Te ~ Ti and of the integration of core confinement and edge compatibility issues
including radiative divertor operation and ELM control. These studies are essential to minimize therisks during initial FPO operation and are not possible during PFPO-1. The integration of coreconfinement and edge compatibility is particularly critical for the ITER mission since, for theconditions required for high Q operation, unmitigated power fluxes are expected to exceed thepower flux capability of the W divertor, excessive W production and core plasma contamination
can lead to the loss of the H-mode and type-I ELMs are expected to cause melting of the plasmafacing components and cause significant W influxes that may also lead to the loss of the H-mode.
Given the importance of demonstrating the compatibility of H-mode operation (with the required
confinement level for high-Q operation) with acceptable edge plasma conditions (i.e. acceptablestationary and transient power fluxes and impurity sources), it is essential that this issue isaddressed thoroughly and as soon as possible in the ITER Research Plan and thus in PFPO-2. This
will allow the development of mitigation strategies to be formulated and refined (e.g. development
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of other high confinement plasma scenarios in ITER alternative to the controlled-ELM H-mode,etc.) in advance of FPO should major issues be identified in the development of the plan described
below.
2.5.5.7.3 Risk to the success of the H-mode studies in PFPO-2
There are five principal issues related to H-mode operation in PFPO-2:
 H-mode access: The basic considerations related to this issue have been outlined in alreadyin section 2.5.4.9 and are further discussed in Appendix B, and an initial assessment of thepower required for H-mode access will have been performed at 1.8 T (for hydrogen andhelium) and at 2.65 T for helium in PFPO-1. Nevertheless, due to the large ripple at the
separatrix for 1.8 T, the low margin for 2.65 T H-mode access in helium plasmas in PFPO-1
and the presence of the TBMs, it is possible that deviations on the expected H-mode
threshold power are identified at this stage that would require a modification of the plan
below. It should also be noted that the application of 3-D fields (particularly, with thealignment required for strong ELM mitigation or suppression) can affect significantly the
H-mode threshold, as originally observed in DIII-D, MAST [Fenstermacher, 2010] and later
in ASDEX-Upgrade [Ryter, 2013]. While an initial assessment of this effect should alreadyhave been performed in PFPO-1, the new flexibility regarding ELM control in PFPO-2 (allpower supplies available allowing a large range of toroidal/poloidal spectra to be applied up
to a maximum current of 90 kAt) may lead to new findings in this area that would require a
modification of the plan below.
 Plasma performance at Q ~ 10 relevant Pinp/PL-H: It is frequently observed that in order toachieve high energy confinement the edge power flow has to exceed the H-mode thresholdby a given margin [Martin, 2008]. Two factors are thought to be at play in determining thisbehaviour one is related to changes in ELM behaviour (type-III to type-I and the need to
avoid long ELM-free periods in devices with high-Z PFCs) and the other to the positive
interaction between core confinement and edge MHD stability leading to an increase of theoverall plasma energy [Garcia, 2015; Urano, 2016] but the details of both remains to beunderstood. Typically, for ITER operation it is assumed that Psep > 1.2 - 1.5PL-H is required
to achieve the energy confinement time corresponding to the ITER-H98(y,2) scaling,although the confinement that will be achieved in hydrogen and helium plasmas is expected
to be lower than in DD/DT. During PFPO-2 the level of available heating will allow
addressing the changes to H-mode plasma confinement at a relevant margin of the edge
power flow over the L-H transition at 1.8T for hydrogen and helium plasmas and at 2.65 T,
most likely only for helium plasmas. If the assessment in PFPO-2 reveals that a significant
margin over the H-mode transition is required to achieve high confinement H-modes in
ITER, this may affect the level to which the characterization of H-modes and thedemonstration of edge–core integration will be performed thus postponing the studies until
FPO and increasing the risk of later operation. In addition, an upgrade of the additional
heating systems should be launched at this stage in order to minimize the risk that ITER will
not achieve the required performance for high-Q operation in DT.
 Conflicting requirements for stationary and ELM power load control and W control:
Although an initial assessment of this issue may already result from experiments during
PFPO-1, it is only during PFPO-2 that the edge plasma parameters during H-modes will
have sufficiently high values to provide the key features of H-mode operation in FPO. Thesefeatures include the low core plasma neutral source due to efficient ionization of neutrals in132

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the divertor, SOL power e-folding lengths in the mm scale and edge power flows of up to
50-60 MW. The control of stationary power fluxes in ITER at these power levels is expected
to require operation with high separatrix densities and, possibly, radiative power dissipation
at the divertor, as discussed in Appendix C. This, in turn, is expected to prevent impurity
penetration through the pedestal by providing impurity exhaust between ELMs [Dux, 2014],
but may also reduce the efficiency of impurity exhaust during ELMs. This, together with thetransient W influxes from the divertor following the ELMs may restrict the range overwhich the integration of all requirements can be achieved at this stage [Dux, 2017].
Depending on the results obtained, this may imply that only suppressed ELM regimes (i.e.
no transient W influxes) are viable to achieve high-Q operation in ITER, or that alternative
ELM-less plasma scenarios to the controlled/suppressed ELMy H-mode need to be exploredat this stage (QH-mode, I-mode, etc.). Performing this assessment as soon as possible in the
ITER experimental program (i.e. in PFPO-2 as considered here) will allow the early
mitigation strategies to prevent this risk materializing in FPO.
 Development of ELM control schemes and plasma-facing component melting avoidance: It
is important that, as ELM control schemes are developed, substantial melting of plasmafacing components is avoided. The strategy developed in PFPO-1 and PFPO-2 is based on
predictions that melting of W divertor monoblock will not occur during uncontrolled ELMs
at 5 MA/1.8 T (including edges) but this may then occur at 7.5 MA/2.65 T (at least fortoroidal edges), as discussed in Appendix C. In view of this, and the considerations on postELM W influxes above, it is important that ELM control is maintained to avoid melting as
the ELM control schemes are developed, particularly because of the fine tuning issues
related to the alignment between the applied 3-D field magnetic perturbation and the edge
magnetic field. This risk is already significantly reduced within the proposed staged
operation plan by the application of vertical plasma position oscillations, which are expected
to be effective to trigger ELMs up to currents of at least 7.5 MA [Gribov, 2015] and by theuse of the four available fuelling pellet injectors as back-up for ELM control by injectingpellets from the LFS and HFS. In this application the expected pellet frequency that can beachieved is up to 64 Hz and should allow WELM ≤ 0.375 MJ over the range of H-modes
explored, if this does not cause excessive fuelling of hydrogen plasmas or dilution of helium
plasmas. If issues are identified at this stage regarding ELM control that would make itquestionable for high-Q plasmas in DT, possible mitigation actions would involve theaddition of two further pellet injector for ELM triggering before FPO operation and or theabove mentioned exploration of alternative plasma scenarios to the controlled/suppressed
ELMy H-mode at this stage.
 Fuelling of He plasmas at 7.5 MA/2.65 T: Fuelling of He plasmas can only be performed bygas fuelling and this is not efficient in ITER to fuel the plasma because of the highionization efficiency of neutrals in the SOL/divertor plasma; pellet fuelling is not possible
due to H contamination of the He H-modes. Integrated simulations that take into account thelimits that divertor detachment imposes to the separatrix density (see Appendix C) andanomalous core particle transport in ITER [Militello-Asp, 2016; Polevoi, 2013], show that
H-mode scenarios exist which allow unrestricted application of NBI both in the access phaseand the stationary phases. If experiments did not reproduce the modelled separatrix density
and core density profiles in these simulations and lower density values were to be achieved(particularly in L-mode), the application of NBI in He plasmas would be more complex andthe He H-mode scenarios more difficult to develop. Possible mitigation strategies consist on:a) reducing the NBI energy and thus power injected, to cope with a lower plasma density133

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with acceptable shine-through loads, and b) triggering the L-H transition with ECRH+ICRF
and then applying the NBI to full power once the density has increased in H-mode.
2.5.5.7.4 Details of experimental plan and deliverables for H-mode scenarios in PFPO-2
The PFPO-2 experimental plan starts with similar H-mode plasmas to those that should be obtained
before the end of PFPO-1 and then expands the H-mode operational range towards 7.5 MA/2.65 T.
Due to the high H-mode threshold of hydrogen plasmas at 2.65 T and the lack of a suitable heating
scheme (for ICRF) or restrictions (NBI shine-through loads) in the application of the baseline
heating systems, the reference research plan in PFPO-2 is to perform comparisons of hydrogen andhelium H-mode plasma behaviour at 1.8 T and the main H-mode scenario development in helium
plasmas at 7.5 MA/2.65 T. Despite this, the plan includes an early assessment of the H-mode
operational range that can be achieved in hydrogen plasmas at 7.5 MA/2.65 T so that the limitations
expected for these plasmas can be corroborated experimentally (high shine-through power loads
requiring high density for NBI and high H-mode threshold) and includes dedicated studies toenlarge this operational range. Following this assessment and studies, it will be possible to decide
whether to conduct the main H-mode scenario development in helium plasmas at 7.5 MA/2.65 T or
if hydrogen plasmas would be a possibility. The latter would obviously advantageous for the
extrapolation of the results obtained in PFPO-2 to FPO but, for the reasons explained above, is notconsidered feasible with the experimental evidence at this stage and the modelling performed for
ITER so far.
2.5.5.7.4.1 H-mode operation to connect with PFPO-1, evaluate/mitigate TBM effects and toextend H-mode operation at 1.8T (40 days)
The main objective of this phase is first to assess the effects of the TBMs on H-mode plasma in thesame plasma conditions as obtained in PFPO-1 at 5 MA/1.8 T, mitigate these effects if sizeable, andthen expand the operational range of 1.8T plasmas, predominantly in hydrogen plasmas to takeadvantage of the increased additional heating power level in this phase.
Based on the results from operation at 1.8T in PFPO-1, the experiments in PFPO-2 will start in theconditions that provide type-I ELMy H-modes with high energy confinement. In principle, in orderto provide a better assessment of the effect of the TBMs on H-mode plasmas, it is preferable toselect plasma shapes whose separatrix-wall distance is as close to the reference ITER shape as
possible while ensuring high energy confinement (if the -1.28% ripple level at the reference
separatrix position turns out to degrade confinement unacceptably) and in hydrogen, if possible (ifnot in helium). The H-mode access experiments with power ramps (20s ramp-up followed by 10 s
stationary) would be repeated in PFPO-2 applying the same heating levels and schemes as in PFPO1. The possible differences regarding H-mode access and sustainment due to the presence of the
TBMs will be documented and, if significant, schemes to mitigate these effects will be developed.
This will be based on the optimization of n = 1 error field correction for H-mode plasmas with theexternal coils and the use of the ELM control coils (now with the full baseline capabilities) to finetune static low-n error field correction and, possibly, to develop active correction schemes.
Once the schemes to mitigate the effect of TBMs have been developed, the program will continue
by expanding the H-mode operational range of 1.8 T H-modes taking advantage of the increased
additional heating power level in this phase and of the full ELM control capabilities with 3-D fields
for both hydrogen and helium plasmas. This will include experiments similar those alreadyperformed in PFPO-1 over a range of currents of 3.5-5.0 MA and will address the scaling of
H-mode confinement, pedestal/SOL plasma parameters with increasing input power and margin
above the H-mode threshold power. The increased input power will also allow the expansion of the134ITR-18-003

operational range to higher plasma densities and to perform an initial evaluation of fuelling
efficiency of gas versus pellet fuelling and particle transport studies to be performed in this phase,
also taking advantage of the almost full baseline diagnostic capabilities. On the basis of theevaluation in Figure 2.5-4, H-mode operation should be possible for both hydrogen and helium
plasmas over the density range (ne = 0.4 – 1.0nGW) with NBI and ECRH heating alone. The
additional 20 MW of ICRF (assuming coupling issues are not important or have been resolved in
PFPO-1) would allow the achievement of helium and hydrogen plasmas at the same level of
Pinput/PL-H ≤ 2 and to push helium plasmas to levels of Pinput/PL-H ≤ 3-4 (if second harmonic
hydrogen minority heating is efficient) at 5 MA/1.8 T. It should be noted that for these plasma
conditions the fast ion populations from NBI and ICRF will be significant as well as the current
driven level [Polevoi, 2012; Maget, 2013]. Therefore the plasmas at the highest power levels may
develop instabilities driven by fast particles or by the driven current profile and thus may not beappropriate for the purpose of the characterization of H-mode plasmas for comparison with plasmas
at larger plasma currents [Polevoi, 2012]. Details on studies related to the use of the H&CD systems
for current drive studies related to long-pulse development and the expected research on fast
particles that can be performed at this stage is discussed in sections 2.5.5.9 and 2.5.5.10.
Together with the expansion of the type-I ELMy H-mode operational space, ELM control
experiments with the full baseline configuration of the ELM control coil power supplies will be
performed. This will include exploration of a range of toroidal mode numbers (most likely fromn = 2 to n = 4) and of poloidal spectra over the range of q95 = 3.0 - 4.3 at 1.8 T (i.e. Ip = 3.5 –5.0 MA), as far as possible. In particular the optimization of the spectra to achieve the same level of
ELM control/suppression to account for changes in q95, <ne>, , and plasma rotation (that can bevaried by adjusting with NBI power level) will be assessed thoroughly. The exploration will also
assess the effects of 3-D fields on H-mode access and identify schemes for the time variation of the
fields so that their effect on H-mode access is minimized while they provide the required level of
ELM control during the stationary H-mode phase. This is an important step that should eventuallyconclude with the demonstration of ELM suppression for 1.8 T plasmas over q95 = 3.0 - 4.3 and thatwill provide an assessment of what will be required to achieve ELM suppression in ITER. As in thecase of PFPO-1, the back-up ELM triggering techniques provided by vertical plasma oscillations
and pellet injection (now with 4 available injectors) will be applied to ensure that the minimum
level of ELM control is maintained as the 3-D fields by the ELM control coils are retuned for eachof the conditions explored, particularly when q95 is varied.
The 1.8 T experiments described above are estimated to require 40 operational days and will beshared between hydrogen and helium plasmas. While priority will be given to hydrogen operation,
it is not possible to predict the operational sharing between the two species. If the most pessimistic
aspects of hydrogen H-mode operation at 1.8 T in ITER materialize (e.g. no viable ICRF coupling
because of high ripple issues, etc.) a significant proportion of the experiments would be performedin helium (at least 25-30 of the 40 operational days are estimated). On the other hand, if this is not
the case, the operation in helium would be restricted to the minimum required to provide:Assessment of the effects of TBMs on H-mode access and confinement and mitigation (ifthe reference plasmas in PFPO-1 were performed in helium), andReference set of H-mode plasma conditions and requirements to control/suppress type-I
ELMs at 5 MA/1.8 T to compare with 7.5 MA/2.65 T helium plasmas in PFPO-2.

In this case, the helium operational period could be reduced to 5-10 days.
Obviously, if this initial assessment of hydrogen H-mode operation at 1.8 T provided, in addition,
sufficient confidence that stationary type-I ELMy H-mode are feasible in hydrogen at135

ITR-18-003

7.5 MA/2.65 T (e.g. lower H-mode threshold than expected and/or lower shine-through loads thancalculated), then the helium operational period would be shortened or fully eliminated at this stage,
depending on whether the PFPO-1 reference H-modes to assess TBM effects were performed inhydrogen or not. If the PFPO-1 reference H-modes are in hydrogen and 7.5 MA/2.65 T stationary
H-mode operation is evaluated to be feasible, then there would be no need to have reference
H-mode and ELM control experiments in helium plasmas at 5 MA/1.8 T to compare with
7.5 MA/2.65 T and to evaluate the TBM effects, as this would be done in hydrogen.
2.5.5.7.4.2 Expansion of H-mode operation towards 7.5 MA/2.65 T in hydrogen plasmas (10 days)
These experiments will assess first H-mode access at 2.65 T and then explore H-mode operation
with increasing power levels and ELM control, in a similar fashion as done for 1.8T during
PFPO-2. As discussed in section 2.5.5.7.1, the expected operational range of hydrogen H-modes at
2.65 T is very narrow both due the lack on an ICRF scheme with high SPA as well as to the high
NBI shine-through loads expected at low density. Therefore, these initial hydrogen experiments will
be focussed towards the determination of the H-mode threshold in hydrogen plasmas at 2.65 T andthe assessment of the possible operational range for stationary H-modes. On the basis of theevaluations in Figure 2.5-4 and Figure 2.5-5, it should be possible to access the H-mode inhydrogen plasmas in ITER for densities in the range of 2.4 – 3.61019 m-3, but the H-mode phasewould be limited to ~ 10 s when all available heating power is applied (33 MW of NBI + 20 MW of
ECRH). To minimize the probability of accidental melting of the W divertor in this exploration, theexperiments will start at 5 MA where melting is not expected to occur, as discussed in Appendix C.
At this current level, <ne> = 2.41019 m-3 corresponds to 0.6nGW and this density is expected to beachieved with pellet fuelling if required [Militello-Asp, 2016; Garzotti, 2016]. The current level
will then be increased in steps of 0.5-1.0 MA towards 7.5 MA, and the temperature of the divertor
during ELMs will be monitored at each step. This scan will provide the scaling of the H-mode
threshold at 2.65 T for hydrogen plasmas to compare with 1.8 T plasmas in PFPO-1 and anevaluation of possible effects of q95 on H-mode access with low ripple plasmas.
It is important to note, as was the case in PFPO-1, that these operational conditions near the L-H
threshold are likely to produce long ELM-free periods that may lead to the termination of the
H-mode by W accumulation with a sudden H-L transition. Similarly, the limitations to NBI heating
associated due to the shine-through loads will restrict the possibilities of providing a soft landing of
the H-mode by ramping down the power slowly in this phase. Therefore, significant effort will be
required to ensure that these plasmas do not terminate abruptly, for which a key ingredient is tomaintain ELM control throughout the H-mode entrance and exit phases, as demonstrated in present
experiments [Köchl, 2016]. This will be provided by the back-up ELM triggering schemes available
for these plasmas (vertical plasma position oscillations and pellet injection); developing an ELM
control scheme with 3-D fields in these marginal H-mode conditions would require a significant
amount of time and is expected to have a low probability of success. Based on the evaluations inAppendix C for the low density range of the proposed scans, the divertor power loads for these Hmode plasma conditions are sufficiently high to consider the use of impurity seeding to reduce themor to carefully limit the plasma time in these low density conditions to avoid excessive W
recrystallization (to less than few hundred seconds). The total time estimated to perform this initialinvestigation of hydrogen H-mode access and operational range is 6 days.
On the basis of the findings of these initial experiments, it would possible to assess whether theoperational space of stationary H-modes at 2.65 T is zero or very narrow, as predicted by theevaluation in Figure 2.5-4 and Figure 2.5-5, or wider. On the basis of this assessment, theexperiments would continue along the steps below (no operational space for pure hydrogen plasmas
136

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at 2.65 T) or move to the program discussed in section 2.5.5.7.4.5 (wide operational space and Hmode characterization and ELM control at 2.65 T to be developed with hydrogen plasmas).
Assuming that the evaluations in Figure 2.5-4 and Figure 2.5-5 are correct, the next step would be
to investigate ways to widen sufficiently the H-mode operational space in predominantly hydrogenplasmas so as to provide a good basis for the experiments in section 2.5.5.7.4.5. Two approaches
can be considered:

One would follow the same line as in PFPO-1 by introducing ~10% helium into thehydrogen plasma to decrease the H-mode threshold, andThe alternative approach explores the possibility of neon seeding the hydrogen plasma todecrease the shine-through loads by increased ionization of the injected neutrals, thus
allowing longer pulses at low plasma densities where the margin to the threshold is largest.

For the first approach, it is assumed that injected helium at levels of ~ 10% will decrease the
H-mode threshold of hydrogen to the intermediate values between hydrogen and helium. Thus, thethreshold for <ne> = 3.51019 m-3 predominantly hydrogen plasma at 2.65 T (allowing a 30-40 s
NBI heating pulse with acceptable shine-through loads) would be in the range of ~ 35.7- 41.5 MW,
depending on the helium H-mode threshold value. While some tests on the viability of thisapproach will have been done in PFPO-1 with 1.8 T plasmas (and possibly earlier in PFPO-2 if thereference pulse for the assessment of effects on TBMs is established with H+10% He), it is possiblethat its full potential cannot be exploited due to high ripple issues at 1.8 T and, therefore, it is
proposed to study it again at 2.65 T. This reduced H-mode threshold in mixed hydrogen-helium
plasmas can thus provide some operational range in type-I ELM H-mode with 53 MW of total
heating, particularly if the H-mode threshold is near the lower end of the estimates. In addition, thepresence of 10% helium in the plasma decreases the shine-through loads approximately inproportion to the helium concentration, so that the duration of 30-40 s full power NBI for
H+10%He plasmas with <ne> = 3.51019 m-3 is conservative.
The other approach relies in increasing the density of impurities in the plasma, as this increases
significantly the effective ionization cross section of the plasma for the fast NBI neutrals (chargeexchange capture by highly ionized impurities) and decreases the shine-through loads for a given
value of <ne>. The drawback of this approach is that core radiative losses increase and may lead toa reduction of the edge power flow (this should be maximized for H-mode access). In addition, thepresence of impurities may raise the H-mode threshold itself [Maggi, 2014; Bourdelle, 2014], as
discussed in Appendix B. Simulations carried out for ITER 7.5 MA/2.65 T plasmas show thatseeding with neon can open a window for operation for hydrogen plasmas with full NBI + ECRH
power during long pulses with acceptable shine-through loads, as shown in Figure 2.5-6. For thispurpose, medium-Z impurities are optimum as they are fully ionized in the ITER core plasma andthus do not contribute to an increase of the radiative losses.

137

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Figure2.5-6:
Requiredneon
concentration for unlimited application offull power NBI heating with acceptable
shine-through loads in hydrogen at
7.5 MA/2.65 T versus normalized (toGreenwald density) plasma density,
showing the density range in which Hmode access is possible.

The total time estimated for the assessment of H-mode access in hydrogen and hydrogen-dominant
plasmas following the approaches outlined above is 10 days. The sharing among the three possible
approaches (pure-H, H+10% He and H+1% Ne) will be determined by the results of theexperiments themselves and by the degree of complexity of the scenario control issues (sudden H-Ltransitions, avoidance of long ELM-free periods, plasma radiative collapse etc.). These have to be
faced when operating near the H-mode threshold and may be more difficult to resolve in some of
the approaches than in others.
At this point of the experimental program a decision will be taken whether to proceed with the
2.65 T H-mode characterization and ELM control demonstration in hydrogen or hydrogen dominant
plasmas (i.e. going directly to program of section 2.5.5.7.4.4), or to perform these experiments inhelium. Given the reduced operational range predicted for hydrogen plasmas and predominantly
hydrogen plasmas, the reference strategy is to switch to helium plasmas at this stage. If this is not
required, the allocated 15 operational days in section 2.5.5.7.4.3 to develop helium H-mode accessand make an initial characterization of H-mode behaviour could be dedicated to the expansion of
the operational range of hydrogen plasmas towards high currents and fields described in section
2.5.5.7.4.5.
2.5.5.7.4.3 Expansion of H-mode operation towards 7.5 MA/2.65 T in helium plasmas (15 days)
If robust H-mode access and long-pulse type-I ELMy H-mode cannot be obtained at 2.65 T inhydrogen or predominantly hydrogen-dominant plasmas, helium plasmas will be studied. The typeof scans will be similar to those described in section 2.5.5.7.4.2, starting with 5 MA/2.65 T plasmas
to avoid melting of the W divertor, and then increasing the current towards 7.5 MA. Given themuch lower H-mode threshold, lower shine-through loads and the wider flexibility in heating mix(NBI, ICRF and ECRH) for helium than is predicted for hydrogen plasmas, the range of plasma
densities that can be explored will be much wider than in hydrogen (in principle, from the minimum
density determined by mode-locking up to <ne> ~61019 m-3 = nGW-7.5MA). In practice, theachievable <ne> range in H-mode is expected to be limited to relatively low values of <ne>/nGW
due to the detachment of the divertor for He plasma, which limits the separatrix density to138

ITR-18-003

2.0 - 2.51019 m-3 for PSOL = 40 - 60 MW, as shown in Appendix C, due to the lack of pellet
fuelling for these plasmas. Integrated simulations of ITER scenarios show that even with thisrestriction from divertor detachment on the separatrix density, and the possibility of gas fuelling
only, stationary H-modes with <ne> ≥ 0.5nGW should be achievable for 7.5 MA/2.65 T in He
plasmas with the standard assumptions regarding particle transport in ITER [Polevoi, 2013;
Militello-Asp, 2016].
Operation in He H-modes should thus allow studies of the q95 and density dependence of the power
threshold in low ripple conditions (2.65 T) which is not possible in hydrogen plasmas. Otherwise
the same issues regarding the control of the plasma in these experiments will need to be faced as forhydrogen plasmas (e.g. avoidance of long ELM-free periods, etc.), but may be quantitatively
different. For example, W sputtering by helium will be larger than for hydrogen, due both to thehigher mass and to the higher divertor plasma temperatures (for the same neutral pressures), as
shown in Appendix C. This may restrict the minimum density that can be explored and require amore careful control of ELM-free periods. On the other hand, the ELM power pulses are expected
to be longer in helium than in hydrogen due to the longer ion transit time [Loarte, 2003; Eich,
2003], thus reducing the accidental risk of melting of the W divertor in these experiments.
Similarly, the wider possibilities for additional heating in helium will ease significantly the controlof the plasma in the H-mode access and exit phases, thus decreasing the risk of disruptions in these
helium experiments compared to that in hydrogen.
Depending on the range of parameters over which helium H-modes have been characterized in
PFPO-2, additional experiments on H-mode access at 1.8 T along the lines of those discussed in
section 2.5.5.7.4.1 may be also required at this stage to obtain a matching set of conditions at 1.8 T
and 2.65 T (for example, with the same toroidal ripple level).
The foreseen operational time required for these helium access experiments is 15 days. This islonger than for hydrogen in view of the wider range of parameters to be explored at 2.65 T in
helium compared to hydrogen and the possible need to perform additional experiments at 1.8 T inorder to obtain a matching set of conditions at the two values of the toroidal field.
2.5.5.7.4.4 Characterization of H-mode operation up to 7.5 MA/2.65 T including ELM control andresolution of scenario integration issues (40 days)
The experiments described in section 2.5.5.7.4.2 or section 2.5.5.7.4.3 will provide an assessment of
the power requirements for H-mode access and post-transition H-mode plasma evolution at 2.65 T
and will determine which main ion species (hydrogen or helium) will be used in the further
characterization of stationary H-modes and ELM control. During this follow-up phase, theexperiments will follow a similar strategy as that foreseen for 1.8 T in section 2.5.5.7.4.1, but with amuch wider scope. The objective of this phase of experiments is to provide a fully integrated
H-mode scenario at 7.5 MA/2.65 T which can then be used as the starting point for H-mode
experiments in DD plasmas during FPO.
The experiments will start at 5 MA/2.65 T, where melting of the divertor target during uncontrolled
ELMs is not expected, and gradually increase firstly the heating power level and duration beyond
the initial ~10 s stationary level demonstrated in the experiments of section 2.5.5.7.4.2 or section
2.5.5.7.4.3 towards ~50 s of stationary heating power. For each power level the density will then be
increased and this will allow the characterization of the changes to pedestal and SOL plasmas andoverall H-mode confinement to be performed. Depending on whether the experiments are
performed in hydrogen or helium, a comparison of gas fuelling and pellet fuelling will be
performed (only possible for hydrogen) at each level of the heating power. The foreseen level of139

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heating for each density level will be between ~ 20% above the measured L-H threshold up to the
maximum heating power allowed by shine-through loads (< 73 MW at 5 MA).
ELM control with RMP fields will be developed taking advantage of the full baseline set of power
supplies and building up on the experience developed at section 2.5.5.7.4.1 for 1.8 T plasmas over arange of q95 = 3.0 – 4.3. The effects of the RMP fields on ELM control/suppression, H-mode
confinement, impurity exhaust and fast particle losses will be determined for a range of heating
mixes with a varying level of NBI and optimized (both for ELM control optimization versus
confinement degradation and core W accumulation avoidance) while ensuring that fast particle
power fluxes on plasma-facing components are acceptable. This will allow the development of
RMP optimization schemes with varying levels of toroidal plasma rotation for each density level
and/or to identify the level of toroidal plasma rotation/input NBI torque that is required for optimum
confinement for a given level of ELM control. In order to ensure that ELMs remain controlled to
the required level while ELM control schemes with RMPs are developed, the ELM back-up
triggering schemes: vertical plasma oscillations and pellet pacing (if hydrogen plasmas) will beapplied. As these experiments will focus on ELM control with RMPs during the stationary heatedflat-tops, the ELM back-up triggering schemes will be essential to ensure ELM control in the accessand exit phases to stationary H-modes.
Once 5 MA/2.65 T stationary H-modes with ELM control have been developed, the plasma current
level will be increased in steps of 0.5 - 1.0 MA towards 7.5 MA. Continuous monitoring of thedivertor temperature during ELMs will be performed to ensure that no melting is caused byuncontrolled ELMs; this is expected to occur at the monoblock edges for currents of ~7.5 MA[Gunn, 2017]. At each step, various levels of heating power and power mixes will be applied andthe density will also be varied (with gas and pellets if in H plasmas, if He then only gas). The
optimization of the RMP fields applied for ELM control will be performed for each level of q 95,
plasma parameters (chiefly  for the plasma response effects) and heating mix are varied. This
optimization is potentially very time-consuming and the sequence of steps will need to be carefully
selected on the basis of previous experience in PFPO-1 and PFPO-2 to ensure that it can be
performed with a reasonable investment of time. These experiments will provide an excellent
characterization of H-mode confinement, pedestal characteristics and SOL plasma parameters withincreasing level of plasma current and decreasing recycling neutral source (as the divertor/SOL
ionization becomes more effective with increasing levels of plasma density). Similarly, if theincreasing plasma  and decreasing q95 leads to the triggering of NTMs, schemes tocontrol/suppress them by the use of ECRH/ECCD will be implemented; this together with the needto avoid central W accumulation may restrict the range of heating mixes that can be explored above
a given Ip.
For the lower core neutral sources expected at 7.5 MA, low density gradients are expected to appear
in the pedestal (similar to DT operation). This should lead to profound changes to W impurity
transport as neoclassical screening dominates in the pedestal [Dux, 2104; Dux, 2017; Polevoi,
2016], which can affect the balance of W exhaust between ELMs and at the ELMs. If these effects
are identified at this level of plasma current, dedicated studies to determine the optimization of
RMP fields to provide control of both power fluxes during ELMs and W exhaust will be performed.
In these conditions with dominant pedestal impurity screening, increasing the ELM frequency andreducing WELM may not provide W exhaust from the core plasma on longer timescales, as shownin Figure 2.5-7 (in this modelled case for DT plasmas) [Polevoi, 2016].

140

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Figure 2.5-7: H-mode operating space for Ip/Bt = 15 MA/5.3 T with separatrix PSOL = 100 MW, and for
Ip/Bt = 7.5 MA/2.65 T with PSOL = 50 MW, assuming no W prompt redeposition at the divertor during
ELMs and that average transport during ELMs can be described by a diffusive model (PELM = fELMWELM= 0.2PSOL): (left) ratio of PSOL to PL-H following post-ELM W influx leading to increased core radiation;(right) long timescale W tungsten accumulation factor; a value larger than 1 implies uncontrolled increaseof the W density in the core plasma and termination of the H-mode by an L-H transition.

Together with the expansion of the plasma scenarios towards 7.5 MA not only the transient ELM
power fluxes to plasma-facing components but also the stationary power fluxes and W sputtering
source will increase, as the SOL power width is predicted to scale as Ip-1. To ensure that stationary
power fluxes remain within the ranges required for the W divertor target (see Appendix C) and toprevent core W contamination, additional impurities are expected to be injected to provide
increased divertor radiation and decrease both divertor surface temperature (to prevent
recrystallization, see Appendix C) and plasma temperature. In addition, time-variation of the RMP
perturbation [Loarte, 2014] may need to be applied to avoid the local power fluxes exceeding thelimits in Appendix C. If this is required as part of the plasma scenario development above a given
current/power level, then as part of the experimental program above, the effects of RMPs of highdensity radiative divertor conditions and divertor power fluxes will be characterized at this stage. If
not, dedicated experiments to assess the effects of RMPs on radiative divertor plasmas and power
fluxes in semi-detached divertor conditions and to decrease off-separatrix localized power fluxes bytime variation of the applied 3-D fields will be performed at this stage. This is likely to require aretuning of the applied fields (including the use of cyclic variations in time) to integrate anincreasing range of requirements (ELM control, W control, radiative divertor operation with low
core plasma impurity contamination) which may affect pedestal plasma parameters and core plasma
confinement. Therefore, experiments to re-optimize the applied RMP perturbation and heating-mix
will be performed to identify the best compromise between core and edge integration in theseplasmas which will guide the development of the ELM controlled H-mode plasma scenarios in DT.
Similarly, experiments to test specific aspects of the control of stationary and ELM-caused power
fluxes, such as those related to operation with reduced wall clearance (especially at the top of thefirst wall near the secondary X-point), will be performed at this stage for 7.5 MA/2.65 T plasmas.
The experiments above will provide fully integrated and optimized stationary H-mode plasma
scenarios in terms of plasma confinement, controlled stationary and ELM-caused power fluxes andimpurity exhaust by 3-D fields. The remaining set of experiments will aim at demonstrating the141ITR-18-003

integration of ELM control, power fluxes and impurity exhaust during the H-mode access and exit
phases with 3-D fields, which will be required for DT operation. As mentioned before, in theexperiments above ELM control in these phases is assumed to be provided by the back-up ELM
triggering schemes: vertical plasma oscillation and pellet triggering (if hydrogen plasmas).
To address the control of H-mode access and exit with 3-D fields, two sets of experiments will be
performed:At two levels of Ip (5 and 7.5 MA), 3-D fields will be applied before or just after the L-H
transition and will be switched-off at the H-L transition or afterwards for a suitable range ofplasma densities, levels of input powers and heating mixes determined by the experiments
above. The goal of these experiments is to demonstrate robust access from L-mode to a highconfinement H-mode and controlled H-mode exit with control of ELMs, impurity exhaustand power fluxes to PFCs provided by 3-D fields. Given the expected effect of 3-D fields
optimally aligned for ELM control on the L-H transition [Fenstermacher, 2010; Ryter,2013], these experiments will most likely concentrate on developing strategies to optimize
the time variation of the applied fields in this phase to prevent deleterious effects on the L-H
and H-L transition while providing the ELM control required.



3-D fields will be applied during the current ramp-phases (at 5 MA for a flat-top current of
7.5 MA) in which the L-H and H-L transition will be triggered by changes in additional
heating. The switch-on and off of the fields around the L-H and H-L transitions will follow
the strategies developed in the previous experiments and then the 3-D fields will be varied intime to provide ELM control through the current ramp-up/down H-mode phases in which
q95 will be varying. The experiments will include a range of density and power waveforms
and current ramp-up/down rates for which the 3-D fields will be optimized. To optimize thetuning of the 3-D fields during the q95 ramps the experience developed for ELM control instationary conditions with q95 = 3.0 - 4.5 in 2.65 T H-mode plasmas earlier in thisexperimental phase will be very useful.

Again, as in all the H-mode experiments performed in PFPO-2, the back-up ELM triggering
schemes: vertical plasma oscillation and pellet triggering (if hydrogen plasmas) will be used toprovide the required level of ELM control while the applied 3-D fields are being optimized toprovide ELM control.
The experiments in this PFPO-2 phase are estimated to require 40 operational days and will provide
fully integrated scenarios, including the H-mode access and exit phases, which are optimized interms of plasma confinement, plasma fuelling and controlled stationary and ELM-driven power
fluxes and impurity exhaust provided by 3-D fields. These will be used as template for further
development in DD/DT and towards higher plasma currents in FPO.
2.5.5.7.4.5 Demonstration of long-pulse H-mode operation (6 days)
The last series of experiments in PFPO-2 will perform a demonstration of a long-pulse H-mode
scenario (potentially up to ~ 1000 s) with the purpose of assessing issues related to processes that
may evolve on very long timescales such as those related to changes in the current profile, etc.
These experiments will build on the results of the assessment of the current drive capabilities of the
ITER systems described in section 2.5.5.10, for which synergies with the experiments in section
2.5.5.7.4.4 will be exploited. Assuming that, in the experiments in section 2.5.5.10, the current drive
capabilities of the ITER NBI and ECRH systems are found in agreement with present estimates andthat fast particle-related instabilities do not cause major effects in the plasma, it will be possible to142

ITR-18-003

2

J (MA/m )achieve ~ 1000 s long pulses in 7.5 MA/2.65 T H-modes even without any significant use of the
lifetime of the central solenoid. This is due to the fact that the current drive efficiency of the NBI
for these plasmas is estimated to be ~ 2.5 – 3 MA [Polevoi, 2012; Maget, 2013]. This, together withthe pedestal bootstrap current leads to 60% of the current to be non-inductively driven, as shown in
Figure 2.5-8. This allows a stationary H-mode of 1000 s to be feasible even with a premagnetization
current of 30 kA (compared to the maximum of ~ 45 kA) in the central solenoid for which lifetime
consumption is negligible; Figure 2.5-9 summarizes the achievable flat-top durations for a range ofplasma conditions versus premagnetization current. As described above, these demonstration
experiments will focus on the determination of effects that may need very long timescales todevelop in ITER such as those associated with the relaxation of the current profile which can affectplasma and impurity transport, long timescale evolution of processes related to plasma-wallinteractions, etc., and to develop strategies for their control if required. The total operational time
estimated to be required for these experiments is 6 days. However, at this point in the experimental
program, the H-mode duration is likely to be limited by the pulse duration for which reliable highpower operation of the H&CD systems has been established; the development outlined here
assumes that it has been possible to commission the high power operation of the H&CD systems
required up to the necessary pulse lengths, or that the extension of the pulse length required canreadily be achieved within the allocated experimental time.
1.0

jtot

0.8

jbootstrap

0.6

jNBICD
johm

0.4

0.2

0.0
0.0

0.2

0.4

0.6

0.8

1.0

sqrt(toroidal)

Figure 2.5-8: Current profile for a 7.5 MA/2.65 T He H-mode with 33 MW of NBI and 20 MW of ECRH
and 20 MW of ICRF showing the main components to the total current (bootstrap, NBICD and ohmic).

143

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H-mode33 MW NBICD
40 MW RF

Flat top duration (s)1000

H-modeH= 0.7 - 1.0
core transport
assumptions

H-mode
Q =10

L-mode33 MW NBICD
20 MW RF

100

L-modeL-modene/nGW = 0.2-1.0 20 MW RF
20 kA L-mode
20 kA H-mode
30 kA L-mode
30 kA H-mode
Max.  L-mode

10

L-mode20 MW RF

Ohmic

Max.  H-mode

5

10

15

Ip (MA)

Figure 2.5-9: Flat-top duration versus plasma current level for a range of premagnetization currents andscenarios for each current level.

2.5.5.7.4.6 Option for low ripple hydrogenic H-modes: H-mode in hydrogen at either 3.0 - 3.3 T
and up to 8.5 or 9.5 MA respectively
An option exists for obtaining hydrogen H-modes in regimes with low TF ripple by increasing thevalue of the toroidal field beyond 2.65 T to enable ICRF schemes with SPA based on the three-ion
scheme [Kazakov, 2015] with ~10% He4 and ~1% He3, following its demonstration in Alcator
C-Mod and JET [Kazakov, 2016]. Although the increase of the toroidal field from 2.65 T to 3.0 3.3 T increases the H-mode threshold power by 10 - 20%, the possibility of utilizing ICRF heating
at up to 20 MW and the possible reduction of the H-mode threshold by the presence of ~10% He,
may open a range for predominantly hydrogenic H-mode plasmas at these higher fields. For
example, for a hydrogen plasma with <ne> = 4.51019 m-3 and ~ 10 % He4 at 3.0 T, for which theapplication of NBI at full power is not restricted, the corresponding H-mode threshold could be in
the range of 47 – 55 MW, significantly less than the total available input power of 73 MW. Several
issues need to be assessed before this option can be considered sufficiently viable to be added tothose in sections 2.5.5.7.4.2 and 2.5.5.7.4.4, or, indeed, to replace one of the strategies presented inthose sections. These include the experimental demonstration of the three-ion heating scheme with
the ion species considered for ITER, the confirmation of the effect of quantities in the range of~ 10% He4 in decreasing the hydrogen H-mode threshold and an assessment of the fast ion losses
associated with this scheme. The heating scheme specified produces very high energy He3 ions and,for these fields in ITER, has a heating deposition profile which is well off-axis, so that there might
be significant fast ion losses to the first wall. Preliminary modelling results show that the off-axis
fast He3 tail accelerated by the ICRF is well-confined, both at 3 T and 3.3 T [Joly, 2017], although
the application of 3-D fields for ELM control may modify this. A further reason for concern is thepossible W accumulation in the core with predominantly off-axis heating. This is not considered tobe very critical because the ICRF heating scheme also provides a significant central heating due towave damping on electrons (~30 - 40% of the ICRF power). In addition, the ECRH deposition
profiles for these fields are moderately centrally peaked at 3.0 T in O-mode (at 170 GHz and 104110 GHz), although predominantly off-axis for 3.3 T. H-mode experiments combining central andoff-axis ICRF heating have been performed in Alcator C-Mod without any significant impact of144

ITR-18-003

high-Z (Molybdenum) accumulation [Rice, 2002], indicating that this may not be a problem inthese ITER conditions. However, both the ICRF scheme and the issue of W accumulation in these
ITER H-mode conditions needs further assessment by modelling and experiment before this can be
adopted as one of the main approaches for H-mode operation to be considered for PFPO-2.
2.5.5.7.5 Deliverables for the H-mode scenarios studies in PFPO-2
Consistent with the objectives of the initial H-mode operation in PFPO-1, the deliverables from
H-mode operation in PFPO-2 are:

Quantifying power required for H-mode access and sustainment and characterization of Hmode plasmas in FPO-like conditions: For PFPO-2 the operating scenarios and plannedexperiments are designed to confirm the assessment of the L-H transition power (PLH) for
both hydrogen and helium and the level of margin above this threshold to achieve high Hmode confinement for at least two different toroidal fields. These experiments (particularly
those at 7.5 MA/2.65 T) will provide the first H-mode plasmas in FPO-like conditions (e.g.
low core recycling neutral source) that will allow the development of optimization strategies
for H-mode confinement in ITER (e.g. variations of additional heating mixes). From theseresults it is also expected that the threshold power needed for H-mode access andsustainment in deuterium and DT plasmas can be predicted with confidence. This will
provide the final confirmation in advance of DD/DT operation that the baseline level of
73 MW is appropriate to achieve ITER’s Q = 10 goals and, if not, a quantification of the
upgrade in power required.



ELM control: The ELM control work in PFPO-2 is designed to verify the techniques to tune
the applied RMP spectrum to optimize the control of ELMs for two values of the toroidal
field and several edge safety factors using the full complement of power supplies for theinternal coils. The range of q95 values (3.0 - 4.5) will span the values needed for the
scenarios in the FPO campaigns from 7.5MA/2.65T to high-Q operation including Q=10
(15 MA/5.3 T) and long-pulse Q = 5 (11.2-12.5 MA/5.3 T). The ELM control validation
with RMPs for multiple q95 values and during current ramps is very important for thedetailed planning of H-mode operation in the first FPO campaign at higher field and current,
for which unmitigated ELMs would cause melting of the PFCs. Similarly, the studies on
ELM control with various heating mixes will provide information regarding the role ofheating mixes on ELM control and its optimization for minimum confinement deterioration.
This can provide information regarding possible upgrades of the ITER systems to improve
this optimization (e.g. the 3rd HNB to increase torque input).



Document the effect of the TBMs on H-mode performance and develop schemes for their
mitigation: Plasma pulses produced in the presence of the TBMs in PFPO-2 will match
those established in PFPO-1 without the TBMs, and will allow documentation of the effectsof the TBMs on H-mode performance, albeit at 1.8 T and most likely for the ITER reference
plasma shape that has a high level of ripple at this toroidal field value. Mitigation schemes
for these effects (if the effects are sizeable) will be developed using the error field correction
coils and the ELM control coils at 1.8 T and applied to plasmas at 2.65 T, which have amuch lower of TF ripple.



Initial validation of predictions of SOL power width in ITER: The variation of the plasma
current for both hydrogen and helium H-mode plasmas planned for PFPO-2 will provide
data to validate predictions of the width of the SOL power flux, i.e. that it scales inverselywith current (linearly with poloidal field) and that does depend on the toroidal field. This145

ITR-18-003

data is important for the planning of high power H-mode operation at high currents in FPO,
in particular the need and schemes for divertor power load control with radiative divertor
operation discussed in Appendix C.
Assessment of scenario compatibility issues for ELM control with 3-D fields (Pellet
Fuelling, Impurity exhaust and radiative divertor). The experiments in PFPO-2 will
demonstrate these three key scenario compatibility issues of ELM control with 3-D fields,
although marginally for pellet fuelling if experiments are done in He plasmas. These are
required for the application of this scheme in high-Q plasmas in FPO. The PFPO-2
experiments will provide guidance on how to use the large spectral flexibility of the ITER
ELM control coils for this demonstration.

An initial assessment of long-pulse H-mode operation. The first demonstration of a longpulse (~ up to 1000 s, if allowed by progress in the H&CD commissioning) 7.5 MA/2.65 T
H-mode scenario with ~ 60% driven current would help to identify processes that may
develop over long timescales in ITER affecting plasma confinement/impurity
transport/MHD stability and the schemes required to control them. This would provide
guidance on how to control such phenomena, which will be used for the long H-mode pulses
required for the demonstration of the Q =10 and Q =5 goals in FPO.

2.5.5.7.6 Assessment of diagnostic capabilities for performing H-mode research during PFPO-2
By PFPO-2 most of the ITER diagnostics will already be in final configuration; the only diagnostics
that will become available for FPO are those for additional neutron measurements (not essential forhydrogen and helium plasmas) and for fast particle measurements. The latter are possibly the only
systems which might be needed in PFPO-2 as the level of additional heating increases and, inparticular, the use of NBI heating at high power levels starts. The triggering of fast particle driven
MHD instabilities (see section 2.5.5.10), together with the application of 3-D fields for ELM
control, might lead to increased fast particle losses (especially from the NB-injected ions) for whosecharacterization the additional capabilities of the lost fast particle diagnostics would be useful. It
should be noted, however, that if the fast particle losses in PFPO-2 are significant the local highpower fluxes associated with them will be measured with the IR cameras. Therefore, an initialquantitative evaluation of the magnitude of the fast particle losses will be possible and the highest
risk of such localized losses to plasma operation will be avoided.
2.5.5.8 Scenarios for long-pulse operation
Long-pulse operation is desirable in PFPO-2 (even in PFPO-1, time permitting) for commissioning
all plant operations for FPO requirements. Of principal concern are the magnetic sensors and theheating and current drive systems. In addition, long pulses facilitate verification of the current drive
efficiencies of the various H&CD systems as these systems and the required diagnostics become
available.
The ITER Research Plan includes two basic scenarios for long-pulse operation in PFPO-2 both at
7.5 MA/2.65 T. The first is a hydrogen L-mode scenario at 7.5 MA/2.65 T with reduced NBI
energy/power and ECRH for retention studies (with D spiking, see section 2.5.5.12 and AppendixA) and a flat-top length of ~100 s. The second is an H-mode scenario (in hydrogen or helium) at
7.5 MA/2.65 T, with 33 MW of NBI and 20 – 40 MW of ECRH+ICRF, having a maximum
potential flat-top length of ~ 1000 s. Both scenarios can be performed with 30 kA premagnetizationand thus without significant consumption of the CS fatigue lifetime. Additional, specific issues
related to the development of long-pulse scenarios Q ~ 5 scenarios for FPO are also described in146ITR-18-003

section 2.5.5.10 within PFPO-2 although many of them do not necessarily require long pulses. Dueto the availability of diagnostics (e.g. MSE) and H&CD (e.g. HNB and full ICRF in addition to
ECRH), it is more appropriate to perform these studies and long-pulse H-mode operation in PFPO-2
than in PFPO-1.
Although not specifically mentioned in the Research Plan, long-pulse scenarios are also possible in
PFPO-1 and could be used for ECRH and ICRF commissioning, diagnostic commissioning, etc.
Given the relatively narrow operational space for H-mode operation in PFPO-1 (see section 2.5.4.9)
and the reduced set of diagnostics at this stage (see Appendix H), it would be more effective toperform such long discharges in hydrogen L-modes. The achievable flat-top durations for 5 MA
plasmas with 20 MW of RF heating are in the range 200 – 700 s, depending on the value of theplasma density (<ne> = 0.2 – 1.0nGW - see Figure 2.5-9) for 30 kA premagnetization and thus nosignificant CS lifetime consumption. These L-mode plasmas would be a natural extension of thecommissioning activities at 1.8 T discussed in section 2.5.4.5 if the reliability of the ECRH andICRF systems is sufficient to attempt this long-pulse operation at this stage.
2.5.5.9 Demonstrate critical system performance (including heat loads)
2.5.5.9.1 Key elements of critical system performance
Experimental operation during the FPO phase will rapidly lead to activation of the tokamak
structures, implying that maintenance, repair and upgrade activities will require remote handlingand are likely to be time consuming. To minimize in-vessel interventions, it is therefore essential
for all operational systems to demonstrate that they are capable of operating as close to theirperformance and pulse length specifications as is feasible during PFPO-2. Due to limitations
imposed by flux consumption, achieving long-pulse operation of up to 400 s, or longer, in PFPO-2
is likely to require operating at modest plasma currents of ~7.5 MA (e.g. Figure 2.5-9). The
program would likely involve both L- and H-mode plasmas, according to the system test involved,
but, as discussed below, certain tests will benefit from H-mode operation. Further optimization ofplasma scenarios may therefore be required to allow the exploitation of the maximum baseline
power capability under relevant conditions, which provide the most demanding tests of thetokamak, auxiliary and plant systems. The PCS must also demonstrate as much of its advanced
control capabilities as possible in H/He plasmas to ensure that any modifications required of theactuators and diagnostic systems can be carried out prior to FPO. Similarly, critical plasma-related
interlock and safety functions within CIS and CSS must be fully commissioned during PFPO-2. The
pellet and gas fuelling systems must demonstrate adequate fuelling capability up to 15 MA
operation, and so some elements of this program can be addressed within the 15 MA/5.3 T
development discussed in section 2.5.5.11.
Once all H&CD systems have been commissioned to their rated power levels, operation at currents
of at least 7.5 MA has been developed and reliable H-mode operation with acceptable ELM control
has been established in PFPO-2, the long-pulse, heat management capabilities of the device will betested under quasi-stationary long-pulse conditions. The pulse duration of high power plasmas will
be progressively extended to the maximum allowed by the available poloidal flux. Flat-top pulse
lengths exceeding 100 s are possible (e.g. Figure 2.5-9) according to the experimental time
available for commissioning of the H&CD, fuelling, diagnostics, control systems, etc. up to longer
pulse lengths. To reach divertor heat flux levels approaching those expected in the DT phase ofoperation, maximum available power will be required in plasmas at the maximum current and field
at which satisfactory H-mode operation can be sustained (see Appendix C). Monitoring of the heat
handling capabilities of all critical systems (e.g., H&CD systems, in-vessel components, plasma147

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facing surfaces in the divertor) will be essential. Given that power densities are expected to be
beyond those previously investigated in an actively cooled tokamak divertor, it is anticipated that
this will be a demanding step, requiring, perhaps, several months, and longer if repairs or
modifications are necessary. Operation in H-mode is preferred, since this will produce the mostchallenging combination of quasi-stationary and transient heat fluxes on the divertor, and feedback
control of the edge radiation through impurity injection will be essential. These experiments will
provide insight into the range of heat loads that may be expected in the future, verify the capability
to measure these heat loads in real time and to take appropriate control actions on a routine basis. In
addition, the output from this phase will provide valuable data for the assessment of divertor codes.
With longer pulses, the heating of all components and structures inside and attached to the vacuum
vessel will require careful monitoring.
The Torus Cooling Water System (TCWS) and cryoplant must also demonstrate the capability ofmaintaining sufficient cooling within prescribed limits at up to 73 MW of auxiliary heating power,both at high (preferably full) plasma current and for pulse lengths potentially up to several hundred
seconds at lower current levels. The cooling capability must be demonstrated with sufficient margin
to give confidence in the ability of these systems to handle the additional expected heat loads due tofusion power in the subsequent FPO phase. Indeed, a related issue is that the AC heat loads in the
Magnet system must be characterized (and be within acceptable limits) to provide a basis for the
extrapolation of plasma scenario demands to the Q = 10 and Q = 5 scenarios foreseen in DT
operation. The combined heating and coil power supply systems must keep (pulsed) electrical grid
demand within the 500 MW site limit and avoid excessive fluctuations on the grid for the mostdemanding emergency termination scenarios.
2.5.5.9.2 Critical system performance deliverables in PFPO-2
The deliverables of critical system performance during PFPO-2 include:








ECRH, ICRF, and NB heating systems must demonstrate full power with pulse lengths
limited only by the available experimental time for commissioning;
Gas and pellet fuelling systems must demonstrate full fuelling capability for 15 MA
plasmas;Demonstration of satisfactory heat load control on first wall and divertor to full auxiliary
heating power capability;
Radiation and impurity profile control demonstrated up to full auxiliary heating power andavailable pulse length;
TCWS and the cryoplant must demonstrate sufficient cooling capability at full auxiliary
heating power and available pulse length to extrapolate to full fusion burn conditions;
All systems must demonstrate that they operate within the 500 MW site power limit without
excessive fluctuations on the grid up to the full auxiliary heating power for the mostdemanding emergency termination scenarios;
CIS and CSS plasma-related interlock and safety functions must be fully commissioned.

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2.5.5.10 Initial studies of current drive efficiency, target q-profile formation and fast particle
physics
2.5.5.10.1 Assessment of H&CD and diagnostic capabilities for performing current drive studies
during PFPO-2
The primary goal of this research in PFPO-2 is to start the demonstration of the essential
components for the development required to achieve the two long-pulse high-Q ITER goals. These
are a stable long-pulse (target 1000 s, high non-inductive fraction fNI ~ 0.6, Q ~ 5) and fully noninductive steady-state pulse (target 3000 s, fNI = 1.0, highest achievable Q, aiming at Q ~ 5)
[Gormezano, 2007]. During these preliminary exploration stages in H/He the research willconcentrate on the assessment of non-inductive current drive efficiencies by various auxiliary(ECRH, ICRF, NBI) and plasma (bootstrap) means, and on the deployment of advanced real-time
control techniques for maintaining plasma stability () and desired current and kinetic profiles.
The experiments will be focussed on the range of q95 = 3.5 to 6. Operation at q95 = 4 will target
long-pulse operation in hybrid discharges [Luce, 2003; Staebler. 2005; Joffrin, 2005; Hobirk, 2012],
while steady-state operations (fully non-inductive) will be targeted by operation at q95 = 5 [Kamada,
1994; Söldner, 1999; Wolf 2003]. The capabilities of the baseline H&CD systems (33 MW of
HNB, 20 MW of ICRF, and 20 MW of ECRH/ECCD) and of the diagnostics (in particular thoserelated to the determination of the current profile shape, such as polarimetry and MSE) alreadyavailable by this PFPO-2 stage will be used for this research. The overarching objective of the
PFPO-2 experiments is to assess current drive capabilities of the baseline H&CD systems in low-L- and H-modes (He or H, although H is preferable) at q95 = 4 and q95 = 5. In order to allow thesestudies to be performed in H-mode plasmas, a toroidal field of 2.65 T is selected (see section
2.5.5.7) and, thus, the plasma scenarios to be explored will range over 4.5 - 5.6 MA/2.65 T. Fromthe evaluation of the operational space in hydrogen H-modes with the baseline heating schemes at
2.65 T in section 2.5.5.7.1, it is expected that most of the experiments in stationary H-modes will be
performed in He. However, if the approaches discussed in section 2.5.5.7.4.2 are successful in
expanding the hydrogen H-mode operational space, it might be possible to perform those carriedout at 5.6 MA/2.65 T, q95 = 4 in hydrogen plasmas.
In order to accomplish the goals of this phase, it is firstly necessary to have the full baseline H&CD
systems available. In addition to these systems, MSE, polarimetry and magnetic equilibrium
calculations incorporating kinetic profiles are required in order to determine the current profiles
produced by these systems. These capabilities should be available by PFPO-2 as well, but it isimportant that these have reached sufficient maturity for their application to support the executionof the experiments. An additional requirement for this phase of operation includes the availability ofreal-time control algorithms for plasma formation, maintaining stability and H&CD actuator control
for the production of desired target current profiles. The experiments should have stationary plasma
conditions in order to produce relaxed j(r) profiles. This will require density and ELM control
systems to be fully operational and demonstrated for H-mode plasmas, which is foreseen in the
PFPO-2 plan (see section 2.5.5.7), as well as the use of ECRH/ECCD for NTM control in thesescenarios, if required.
As discussed in Appendix E, the flexibility of the baseline H&CD systems is not expected to be
sufficient for high-Q operation in steady-state DT plasmas. Therefore, it is anticipated that a major
activity during this initial scenario development will be to evaluate the capability of these systems
to produce steady-state operation and to establish a firm physics basis for decisions on potential
upgrades of these systems. It is anticipated that, even if fully non-inductive operation cannot be149

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achieved with baseline heating and current drive systems at this stage, the required q-profile can be
maintained for sufficient duration (t < tRESISTIVE) to allow assessments of the associated plasma
performance, fuelling requirements, control needs, divertor compatibility and MHD plasma stability
in PFPO-2. The results of these studies will provide first experimental evidence for the choice ofbaseline heating and current drive upgrades to be implemented to ensure that the Q ~ 5 long-pulse
and steady-state goals of ITER can be achieved in FPO. From the evaluations in Appendix E, these
are likely to be a 3rd HNB (to provide more current drive and input torque) or/and an additional
20 MW of ECRH/ECCD (to provide more heating and j(r) control capability).
To a large degree in parallel with this research, but also with the program of H-mode studies(section 2.5.5.7) and L-mode studies (section 2.5.5.11) in PFPO-2, investigations of fast particle
populations and their effects on MHD stability will be launched at this stage. This will require thecommissioning of the relevant diagnostics which will be available in PFPO-2 or before. Possibly
some additional points along the experimental scans already included in the main program activities
will be required to characterize or control specific fast particle physics processes. One of the key
physics issues to be assessed in this phase is whether the NBI fast ion confinement can be describedby classical processes, or whether it is anomalous due to the presence of Alfvén Eigenmodes orother MHD processes. To this end, it is important that as many fast ion diagnostics as possible arefunctional for these experiments and, therefore, it would be beneficial if the installation andcommissioning of the fast ion loss detector and the vertical and radial gamma ray spectrometers
were accelerated so that they are available by PFPO-2; presently these are foreseen for FPO.
The overall distribution of the experimental time among the various issues to be address in thisperiod is given below:
Scans of the ECRH/ECCD power deposition at constant density in the current flattop of
stationary H-modes to evaluate heating and current drive: 4 days
Scan of NBI beam sources (on-axis vs. off-axis) in the current flat-top of stationary
H-modes and the assessment of NBI fast ion confinement: 2 daysAttaining target j(r) at the end of the current ramp-up phase (L-mode) using real-time
control algorithm and extension ~10 s: 10 days

2.5.5.10.2 Risk to the initial studies of current drive efficiency, target q-profile formation and fast
particle physics
The major risks to this campaign are associated with technical difficulties related to the available
heating and current drive systems and/or their integrated control, insufficient fast particle
diagnostics and uncontrolled behaviour of plasmas in ITER, namely: Heating and current drive and/or control schemes cannot provide the required capabilities
to produce and/or sustain the target j(r) required for the experiments: Depending onwhether these issues are related to specific features of the H/He plasmas or not the
mitigation strategy will vary. To mitigate this risk more experimental time could beallocated to refine the control schemes in PFPO-2. If the inefficiency in current drive orcontrol results from the specific features of these H/He plasmas (e.g. Te is too low for
efficiency current drive, density control in He H-modes does not allow a proper evaluationof CD efficiency, etc.), which are not expected to occur in D/DT plasmas, then thisoperational period would be shortened as it is not likely to provide useful information. On
the other hand, if the inefficiency in current drive is expected to affect D/DT plasmas, it will
be necessary to ensure that this is mitigated by the beginning of FPO, for which an upgradeof the baseline heating and current drive systems would be required.
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 Plasma density and temperature cannot be sufficiently well controlled to assess current
drive efficiency: This is a possible risk by this stage, although not likely due to the extensive
H/He L-mode and H-mode program in PFPO-2 that will have been performed before theseexperiments. However, if this becomes an issue, additional dedicated experiments to refine
the level of kinetic plasma control will be required.
 Deficiencies to diagnose specific plasma parameters may affect the execution orinterpretation of the experimental results: Possible issues are related to the measurement of
the current profile and of fast particles characteristics. The former are well covered by theavailable diagnostics in PFPO-2 (which are the same as those in FPO for this) but it may befound that these are not sufficiently accurate. In this case an upgrade of the back-end
diagnostics could be required before proceeding to long-pulse experiments in FPO.
Regarding fast particles, the capabilities to measure them are more restricted in PFPO-2 thanin FPO due to the lack of neutrons. Fast particle measurements are mostly based on the
diagnosis of their effects on the plasma (fast magnetics, SXR, ECE, interferometry andpolarimetry) and the measurements of their losses on the first wall and divertor (infrared), as
described in Appendix H. Otherwise, the only other available diagnostic at this stage is theneutral particle analyzer that would provide a line integrated measurement. Therefore it isessential that the fast ion charge exchange diagnostic upgrade foreseen for PFPO-2 inAppendix H is implemented and brought to its fully functional capabilities at this stage. In
addition, it would be highly desirable to advance the fast ion loss detector and gamma ray
spectrometers to be functional in PFPO-2; they are presently foreseen for FPO.
2.5.5.10.3 Detailed experimental plan for the initial studies of current drive efficiency, target qprofile formation and fast particle physics (16 days)
Initial research towards producing high non-inductive current fraction and ultimately fully noninductive discharges with significant fusion power production will start during the non-active
PFPO-2 campaign. An efficient progression to the development of non-inductive steady-state
plasmas would require a detailed assessment of the capabilities of the heating and current drive
systems and the associated operating space during the H and D phases. In particular, the current
drive in low- L-mode and H-mode plasmas (H or He) will be assessed at q95 = 4 (5.6 MA/2.6 5T)
and q95 = 5 (4.5 MA/2.65 T). Specific experimental activities and objectives of this research
include: Commissioning of the various H&CD systems for their respective current drive missions
and identification of the restriction for their application, if this has not been done up to thisstage (shine-through for NBI, incomplete absorption of ECRH, etc.).
 Commissioning of start-up scenarios consistent with these limitations for 5.6 MA/2.65 T and
4.5 MA/2.65 T and of the systems required for q-profile control during the ramp-up phase. It
should be noted that if MSE is required for the q-profile control scheme in the ramp-up
phase, the HNB energy should be substantially reduced to allow injection in the ramp-up or
short beam blips will be required.
 Producing the target current profiles required for hybrid and non-inductive scenarios at endof the ramp-up phase. This includes development for partially non-inductive hybrid
discharges (q95 = 4) as well as for ultimately fully non-inductive, steady state scenarios(q95 = 5) in H-mode. The stability and confinement properties of the plasmas obtained will
be assessed for a range of current/pressure/(and if NBI can be applied) rotation profiles.

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 Assessing the ability to sustain the target current profile obtained at the end of the ramp-up
during the current flat-top beyond the confinement time scale, e.g. up to 10 s. This will
require the tuning of the H&CD capabilities as well as of the plasma density to maintain thiscurrent profile control, in particular if H-mode conditions are attempted at the start of theflat-top.
 Determining the EC current drive efficiency through scans of the EC deposition location bymeasuring the resulting j(r) for q95 = 4 and = 5 stationary L-mode and H-mode plasmas anda range of densities and temperatures (controlled by the application of the other heating
schemes NBI and ICRF).
 Determining the NB current drive efficiency through a scan of the NB deposition location
by measuring the resulting j(r) for q95 = 4 and q95 = 5 stationary H-mode plasmas and arange of densities and temperatures. The specific scans considered in this period rely on theuse of one HNB on-axis and one off-axis. The scans will include assessing the heating andcurrent drive of both beams separately as well as simultaneously in H-mode plasmas. This
will include scans with similar total input power (i.e. ICRF or ECRH should be applied toreplace the reduced HNB power when one HNB is not used) and others in which the total
power is scanned.
 Validating the models that describe current drive. For inductive operation, this will include
assessing Spitzer versus neoclassical conductivity models. For non-inductive operation,neoclassical bootstrap models will be assessed as well as whether neutral beam ion
confinement is classical or anomalous due to the presence of Alfvén Eigenmodes or othermodes. The assessment of ICRF current drive will only be performed for He plasmas forwhich an effective heating scheme at 2.65 T is available.
 Determining the experimental behaviour of fast ions and validating the models for fast ion
behaviour resulting from the fast ion populations arising from all heating systems. In
particular, if MHD instabilities are observed, determining the nature of the instabilities
observed (e.g. global TAE modes in the outer region of the plasma) [Pinches, 2015].
 If Alfvén Eigenmodes (AE) become unstable, investigating the effects of the various heating
and current drive systems on this instability and of the capabilities of these systems to affectAE instability by local modification of plasma parameters (since heating influences theLandau damping experienced by these modes and also changes the spectrum ofcompressional and Alfvénic modes) [Lauber, 2015]. If suitable actuators are established for
AE control then development and refinement of AE control techniques and associatedcontrollers could start at this stage.
 Determining and validating the effects of ELM control coils on fast ion losses for plasmas
with q-profiles as required for hybrid and steady-state plasmas. This would becomplementary to the experiments in PFPO-2 for the inductive scenario in section 2.5.5.7.
It is important to note that many of the activities above may involve the same experimental
conditions and do not require separate experiments. For instance, the experiments to determine theefficiency of ECRH/ECCD in H-modes with NBI (+ ICRF) heating are also likely to provide
information regarding the possible use of ECRH/ECCD to affect AE stability if AEs are driven
unstable by the NBI. In addition, other experiments carried out in PFPO-2 inductive H-mode
scenarios (see section 2.5.5.7) may have addressed some of the issues above already or will provide
guidance for experiments in this phase, thus shortening the required experimental time. For
example, the measurements of NB current drive efficiency may be already available from similar
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experiments where the heating mix will be varied to identify effects on H-mode performance andELM control. Similarly, the effects of ELM control coils on fast ion losses will have beencharacterized as part of the ELM control experiments within a range of q95 = 3 - 6, etc..
2.5.5.10.4 Deliverables for the initial studies of current drive efficiency, target q-profile formationand fast particle physics
The main deliverables from this phase are:

Initial demonstration of fully integrated scenarios with q95 = 4 and 5 and j(r) profiles
required for advanced H-mode scenarios sustained over ~10 s, including the formation ofsuitable j(r) profiles in the ramp-up phase;Definition of control algorithms to achieve the target j(r) profiles for q95 = 4 and 5 and tosustain it in L-mode and initial attempt at H-mode plasmas;

Assessment of NBCD in H-mode plasma conditions for q95 = 4 and 5 and j(r) profiles
required for advanced H-mode scenarios, which can substantiate the need for a 3rd HNB toachieve the Q ~ 5 long-pulse goal;Assessment of ECCD in L-mode and H-mode for q95 = 4 and 5 and j(r) profiles required foradvanced H-mode scenarios, which can substantiate the need for additional ECRH/ECCD
power to achieve the Q ~ 5 long-pulse goal;Assessment of j(r) formation by application of ECRH/ECCD in the ramp-up for q95 = 4 and
5 scenarios foreseen for advanced H-modes and of its flattop sustainment with
ECRH/ECCD and other heating schemes;Assessment of the level of ECRH/ECCD required for NTM control for q95 = 4 and 5 and j(r)profiles required for advanced H-mode scenarios, which can substantiate the need for
additional ECRH/ECCD to achieve the Q ~ 5 long-pulse goal;Assessment of the level of ECRH and ICRF needed for W control for q95 = 4 and 5 and j(r)profiles required for advanced H-mode scenarios. This is expected to be different in thesescenarios as core MHD is likely to change (e.eg less frequent or not sawteeth) and cansubstantiate the need for additional ECRH/ECCD to achieve the Q ~ 5 long-pulse goal;Assessment of the fast particle behaviour and impact on MHD stability for q95 = 4 and 5 andj(r) profiles required for advanced H-mode scenarios and identification of possible actuators
to control this behaviour;

Validation of models required to design hybrid and steady-state scenarios to be further
developed in FPO.

2.5.5.11 Plasma operation at full technical performance (15 MA/5.3 T)
Once plasma operation, plasma control and, especially, the disruption management/mitigation
capabilities are sufficiently reliable, the research program will address the extension of theoperational space to full current (Ip = 15 MA). In preparation of the Fusion Power Operation (FPO),
this phase (PFPO-2) will provide the first experience of high current operation in ITER. The main
aim of this phase is to develop the basis of high current plasma scenarios and to provide thetechnical demonstration that ITER can operate at its baseline design parameters in terms of plasma
current and toroidal field. It will assess the reliability of magnetic control, in particular verticalstability control, the robustness of the plasma current ramp-up and ramp-down, the efficiency of153

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plasma fuelling with gas and pellets, as well as developing schemes for divertor heat flux control
and W production-transport control.
Commissioning of the heating systems and heating schemes will be an integral part of this activity
as well as further optimization of the disruption mitigation scheme. Deliberate unmitigateddisruptions at high current, in particular at 15 MA, are not required to proceed to the FPO phase:
disruption loads will be extrapolated to 15 MA from the disruption tests at lower plasma currents,using appropriate disruption models (as described in section 2.5.5.5). Unmitigated disruptions at
plasma currents above Ip = 7.5 MA should be avoided as far as possible to ensure a fast progress
towards the FPO. This requires a well-developed disruption mitigation function (mitigation scheme,
mitigation trigger) and effective disruption avoidance schemes as well as a targeted operational
campaign to demonstrate high current operation within the proposed operational time. This isestimated to amount to 25 operational days and to consist of several Ip/Bt steps along anapproximately constant q95 path.
The extension to high current in PFPO-2 will be developed via hydrogen L-mode operation, in viewof the limitations on H-mode operation at plasma currents and fields above 7.5 MA/2.65 T both for
H and He plasmas (see Appendix B and section 2.5.5.7.1). The target flat-top duration at fullcurrent and field for hydrogen L-mode operation is ~ 50 s. Operation at 15 MA/5.3 T requires thefull capabilities of the central solenoid and optimization of the current ramp-up/down together withthe use of additional heating. Scenario simulations of 15 MA/5.3 T plasmas shown in Figure 2.5-9
indicate that, for ohmic plasmas, a flat-top of 10 s can be achieved, while 20 MW of RF heating(ECRH or ICRF) will allow the extension of the flat-top to 50 s. For lower values of the plasma
current, similar or longer flat-top lengths can be achieved in heated L-mode plasmas but in this caseat lower premagnetization currents for the CS (≤ 30 kA) and, thus, with negligible fatigue lifetime
consumption.
2.5.5.11.1 Requirements
Before proceeding to expand the operational space towards 15 MA/5.3 T in hydrogen L-mode
plasmas, the following scenarios/technical demonstrations will have been performed:
Operations in hydrogen L-mode plasmas with 7.5 MA/2.65 T and additional heating power
levels of ~ 20 MW will have been demonstrated, in particular with on-axis and off-axis
ECRH heating, as well as for 7.5 MA/5.3 T plasmas. The off-axis case is essential toaddress the risk of the foreseen L-mode scenarios at intermediate fields between 2.65 and5.3 T, as, for some of them, ECRH heating will be rather off-axis. Plasmas with off-axis
heating can be prone to W accumulation (although this phenomenon is more likely in
H-modes) and this can significantly increase the risk for disruptions.



Operations with plasma currents above Ip = 7.5 MA will significantly increase the potential
for large electromagnetic disruption forces. Unmitigated disruptions in this research phaseshould be thus avoided as much as possible. Therefore, as mentioned above, this operational
phase can only start when ITER operation has demonstrated a satisfactory capability of
disruption avoidance and when the disruption mitigation function (mitigation scheme,
mitigation trigger) has reached reliable performance.

The heating and current drive systems will have been commissioned to both power anddurations levels to allow high Ip/Bt operation with flat-top durations of at least 50 s, which
can require heating of the plasma during the current ramp-up phase. This is essential toachieve this flat-top duration at 15 MA/5.3 T, as it has been assessed that for ohmic plasmas

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only a 10 s flat-top is achievable, following a 65 s ramp-up, as shown in Figure 2.5-9. For
lower plasma currents, the use of plasma heating allows the achievement of these flat-top
lengths without the need for a fully charged CS at breakdown and, thus, without consuming
CS fatigue lifetime for these experiments.
The execution of this experimental program requires the availability of the full gas injection
system (for hydrogen and impurity fuelling) and of the four hydrogen fuelling pellet
injectors, full axisymmetric PF control capabilities, vertical stability systems VS1 and VS3,etc.



By the beginning of PFPO-2, the ITER diagnostic set will be close to its final configuration(except some neutron and fast particle diagnostics) and, therefore, it is found to beappropriate for the foreseen L-mode operation.

2.5.5.11.2 Risks to L-mode plasma operation at full technical performance (15 MA/5.3 T)
The major risks for this operational phase are:
Issues associated with the development of breakdown scenarios at intermediate fields
between 2.65 T and 5.3 T. This may require additional experimental time for thisdevelopment and if, unsuccessful, the need to adjust the current and toroidal field level forthe intermediate Ip/Bt steps. In the worst possible case, all the intermediate steps would be
performed at Bt = 5.3 T and thus with q95 varying from 6 to 3.



Recurrent W accumulation for plasmas with intermediate fields between 2.65 T and 5.3 T
and thus off-axis ECRH. This will require additional experimental time to develop thesescenarios and if, unsuccessful, the need to adjust the current and toroidal field level for theintermediate Ip/Bt steps. In the worst possible case, all the intermediate steps would be done
at Bt = 5.3 T and thus with q95 varying from 6 to 3.

Occasional failures in the mitigation of disruptions as Ip/Bt are increased. Unmitigateddisruptions above 7.5 MA can potentially produce large electromagnetic forces and haveserious consequences for the plasma-facing components. Such events could slow down theexperimental program towards 15 MA.

2.5.5.11.3 Detailed experimental plan for the expansion of L-mode plasma operation towards
full technical performance (15 MA/5.3 T) (25 days)
This operational phase is estimated to require 25 operational days and in them the plasma current
will be increased in gradual steps from Ip=7.5 to 15 MA. The gradual increase of Ip/Bt allows thebuilding of robust scenarios at one current level before proceeding to the next step and thus
avoiding unexpected problems at higher currents. This requires that at each step it is verified if: a)there are operational limitations which affect their feasibility, e.g. enough available auxiliary poweris available; b) the step is useful; that is, if it provides an appropriate stepping stone for operation at
higher currents; and c) it provides relevant, new information and does not duplicate otherexperiments already done within the Research Plan. Further details of the non-active operations
scenarios can be found in [Kim, 2016].
The time allocated to this operational phase is optimized to achieve the deliverables which arerequired for FPO and special care should be taken in this phase of operations, as unmitigateddisruptions at high currents could delay the start of FPO. The choice of the intermediate Bt steps
requires the optimization of two constraints related to the location of the ECRH resonance: one isrelated to the need to ensure that EC-assisted breakdown remains effective and the other that the155ITR-18-003

ECRH deposition profile does not move to a position so off-axis that is ineffective in preventing W
accumulation. This is an issue that can lead to increase disruptivity and thus to be avoided whenincreasing Ip beyond 7.5 MA. The assessment of this issue for 7.5 MA/2.65 T plasmas with on-axis
and off-axis heating is thus included in PFPO-1.
The detailed steps in the proposed increase of Ip/Bt to 15 MA/5.3 T depend, therefore, on the resultsof this ECRH on- and off-axis assessment and of the commissioning of the VS system. The increaseof Ip/Bt is proposed to be carried out at values of constant q95 = 3 - 4, as this has advantages related
to the avoidance of disruptions, for which forces increase gradually as ~Ip×Bt. Maintaining anapproximately constant q95 from one Ip/Bt step to the next ensures that the plasma MHD stability forthe next higher Ip can be inferred from the previous one. This reduces the risk of disruptions and, inparticular, the need to redevelop the ramp-up/down and heating termination scenarios for each stepof Ip/Bt, which are usually the more critical phases of the discharges in this respect (although thetermination of heating may be less critical for these L-mode discharges than for later ones in
H-mode). The operational steps proposed in this phase are shown in Figure 2.5-10. Point 1 in thedevelopment path lies underneath point 2 - the difference between the two scenarios is the setting of
the poloidal launch angles, such that for point 1 the ECRH deposition is as central as possible, andfor point 2 it is tuned off-axis to match the innermost radial range accessible with ECRH at point 3(the 2nd harmonic resonance moves to the low field side as the field is increased). Points 4-6
correspond to a similar radial range of off-axis heating using the fundamental resonance on the highfield side. For points 7 and 8, the ECRH is deposited centrally again. Note that the centre of the
ECRH heat deposition profile depends on the temperature profile. The optimum values of theintermediate fields must therefore be confirmed by modelling.

Figure 2.5-10: Operational path (in Ip/ne versus BT space) from 7.5 MA/2.65 T hydrogen plasmas towards
15MA/5.3T L-mode operation, indicated by a numbered series of scenarios.

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For each Ip/Bt level, optimization of the current ramp-up/down to minimize flux consumption
compatible with an acceptable level of W concentration will be performed. This will include the use
the various heating systems applicable (mostly of ECRH and ICRF) to heat the plasma in thesephases; NBI can possibly be used but with lower energy/power due to the low densities expectedduring the lower current phases of the ramp-up/down. Similarly, gas fuelling and pellet fuellingscans will be performed. In particular, pellet fuelling in hydrogen L-mode plasmas is likely to beroutinely applied for the unrestricted application of NBI in these hydrogen L-mode plasmas forwhich <ne> ≥ 4.51019 m-3 is required [Militello-Asp, 2016]. These fuelling scans (including gas
versus pellet comparisons and gas+pellet optimization) will be performed also for the flat-top phase
together with additional power scans (including variations of the heating mix). It is expected that,for high power levels (potentially up to 73 MW of power can be coupled to the plasma for the steps
with Bt > 4.5 T) or/and low values of <ne>, the level of power load to the divertor will be significant
and W sputtering may also need to be reduced/controlled. This will require the development ofdivertor load mitigation and W sputtering reduction schemes by access to radiative divertor
conditions with extrinsic impurity seeding in advance of the exploration of these plasma conditions.
The final goal of the experiments will be the demonstration of fully integrated L-mode scenarios interms of W and divertor power load control (including ramp-up/down) by the application of theadditional heating and fuelling systems in ITER, which minimize flux consumption and that are
MHD stable (including VS) up to 15 MA/5.3 T. These will be the basis for the further developmentof the DD and DT corresponding L-mode scenarios on which the H-mode scenarios will be built on.
In addition to the scenario development discussed above, these experiments will provide the firstexperimental assessment of core particle and energy transport and core plasma confinement at highIp/Bt in ITER. Their dependence on plasma density and fuelling schemes will certainly be assessed,
as well as on additional heating power level and mix, within the general scenario development plan
discussed above. For specific conditions, or for the assessment of specific physics effects, dedicatedexperiments will be performed beyond those strictly required for scenario development.
Experiments of this type could be, for example, the evaluation of the effect of torque input on Lmode transport and confinement for pure NBI heating versus ECRH or ICRF or ECRH+ICRF
heating, the evaluation of the effect of more localized power deposition profile, versus wider oneson plasma confinement and transport, etc.
It may turn out that not all the steps 1-8 in Figure 2.5-10 are feasible, as the efficacy of ECRH toprovide both assisted breakdown and W accumulation control may not be sufficient. Depending on
the results of the experiments assessing these issues (preferably done in PFPO-1 as planned, or, if
not possible, then earlier in PFPO-2), two paths are considered to achieve the final goal of thisphase, which is a 15 MA/5.3 T L-mode hydrogen plasma scenario with ~50 s of flat-top. These aredescribed in Table 2-7 for the path assuming that off-axis ECRH heating can meet the required
scenario needs and in Table 2-8 for the last steps of the path (after 7.5MA/2.65T) which relies onmaintaining on-axis ECRH heating, as far as possible, in most of the steps. For the latter it isassumed that control of W accumulation control will be less efficient with off-axis ECRH heating
and thus the number of pulses with use of ECRH off-axis are decreased in order to reduce thenumber of possible disruptions.
In both cases, scenario #5 (in Figure 2.5-10) is deemed not necessary for the hydrogen L-mode
phase. However, this will be developed in deuterium L-mode plasma in preparation for the
deuterium H-mode in the FPO phase, which will follow a more careful approach when increasing
Ip/Bt (as discussed in section 2.6.3.6 and section 2.6.3.3), as not only the magnetic forces increase,
but also the plasma energy is higher, which can lead to localized melting of first wall and divertor
target plasma-facing components during transients.
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Table 2-7 – Path to 15 MA/5.3 T (in hydrogen) with ECRH off-axis (except #7and #8)
#

Ip/Bt

2

Available heating
PECRH (MW)PICRF (MW)PHNB (MW)

7.5 MA/2.65 T

20

010-33*

3

9.5 MA/3.3 T

200> 10**

4

9.5 MA/4.5 T

>13

20

> 10**

6

12.5 MA/4.5 T

>13

20

33

7

12.5 MA/5.3 T

>13

20

33

8

15 MA/5.3 T

>13

20

33

* These NBI power ranges have been calculated for different assumptions of Greenwald
density fraction [Kim, 2016]; see also section 2.5.5.7 (H-mode PFPO-2)].
** NB heating above 10 MW shall be possible for all scenarios with a proper adjustmentof the acceleration voltage and power modulation to limit shine-through losses. These
will determine the level of NBI heating applicable in these cases and remains to beevaluated.

Table 2-8 – Path to 15 MA/5.3 T (in hydrogen) with ECRH on-axis
#

Ip/Bt

4

Available heating
PECRH (MW)PICRF (MW)PHNB (MW)

9.5 MA/4.5 T

>13

20

>10*

6

12.5 MA/4.5 T

>13

20

33

7

12.5 MA/5.3 T

>13

20

33

8

15 MA/5.3 T

>13

20

33

* NB heating above 10 MW shall be possible for all scenarios with a proper adjustmentof the acceleration voltage and power modulation to limit shine-through losses. These
will determine the level of NBI heating applicable in these cases and remains to beevaluated.

2.5.5.11.4 Deliverables for the experimental plan for the expansion of L-mode plasma operation
towards full technical performance (15 MA/5.3 T)
The main deliverables of this operational phase are:

Demonstration of first L-mode operation at full current and field (15MA/5.3T) with ~ 50 s
flat-top full integrated plasma scenarios including plasma position and vertical stability
control, power load control, W accumulation control, etc.

Demonstration of the intermediate Ip/Bt steps to be followed to achieve 15 MA/5.3 T
L-mode plasmas along a q95 = 3 - 4 path, which will form the basis of the further
development of L-mode scenarios in DD and DT (and for the corresponding H-mode
scenarios) in FPO.

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Assessment of the effects of gas fuelling and pellet fuelling and additional heating power
level and mix on plasma transport and confinement in ITER L-mode scenarios up to
15 MA/5.3 T including ramp-up/down phases.

Optimization of flux consumption in L-mode ramp-up/down phases by using the ITER
systems’ flexibly, in particular, regarding the current ramp rate, and for plasma fuelling andheating in these phases.

2.5.5.12 Edge physics and PWI studies in PFPO-2
2.5.5.12.1 Wall conditioning
The efficiency for the removal of medium Z impurities of hydrogen or helium GDC and ICWC, forwhich scenarios will have been developed in PFPO-1, will be further assessed in PFPO-2. Here it isexpected that PFCs will have evolved with operation time and that significant co-deposition of
impurities with Be eroded from the first wall will have occurred. Continuous follow-up of wallconditions and routine monitoring of impurity sources in reference pulses (see Appendix D) will
therefore be pursued. Since the amount of available ICRF power will be increased in PFPO-2,
ECWC is no longer considered (cf. section 2.5.4.10.1.2), though scenarios will have been madeavailable in PFPO-1.
During PFPO-2, the efficiency of ICWC as a method for the control of the T-inventory build-up in
the machine during the FPO phase can be studied in detail. However, isotopic exchange between H
and D with ICWC can only be quantified if D-trace experiments can be performed or if D-ICWC
discharges are allowed. If so, this implies the development of half and full field ICWC scenarios at
f/Bt = 7.0 – 10.5 MHz/T, in H as well as in in D, knowing that D-ICWC will in fine be operated forthe control of the T-inventory in the FPO phase. As in PFPO-1, machine operation time will be
required to optimize isotopic exchange efficiency vs. pressure, power, poloidal field patterns andduty cycle. The benefit of using the 2nd ICRF antenna at a higher frequency (55 MHz), as
demonstrated in ASDEX-Upgrade and JET-ILW may also be assessed. In addition to torus pressure
measurements and RGAs for the gas balance, partial regeneration (to ~100 K) of the cryopumps
will allow the quantity of removed and retained isotopes to be measured using gas chromatography.
2.5.5.12.2 Fuel retention and material migration
2.5.5.12.2.1 Deuterium trace experimentWith the availability of the divertor erosion, dust and tritium monitor diagnostics, PFPO-2 offers
the possibility for detailed validation of material migration and fuel retention modelling. Such an
exercise requires well-defined and diagnosed plasma conditions, detailed information on material
erosion and transport, co-deposition rates and fuel content in the co-deposits. The best approach istherefore to perform a dedicated experiment at the end of PFPO-2, during which the same discharge
is repeated to accumulate incident particle fluence on the wall and divertor and sufficient thicknessof co-deposits. The tritium monitor, a combined laser-induced desorption and lock-in thermographysystem, is designed to measure the fuel content in and thickness of Be co-deposits forming on thebaffle region of the inner divertor. The system requirements indicate a minimum measurable
thickness of 1 m and a measurement range of 1 - 500 m.
To provide more information on the spatial distribution of Be deposition and fuel retention in thedivertor, it is proposed to remove a divertor cassette from the machine during Assembly Phase IV,
immediately following PFPO-2, to allow post-mortem analyses in a way similar to what is159

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performed routinely after a JET-ILW campaign. Preliminary discussions indicate that a new set ofplasma-facing units can be fitted on the extracted cassette which could then be re-installed in themachine for the start of the FPO phase. This would avoid the need to use one of the four divertor
cassette spares. Once removed, samples could be extracted destructively from the plasma-facing
units and various surface analysis techniques deployed to measure the poloidal distribution of the
Be co-deposition.
One important output from this exercise is that information on the density and fuel content of theformed Be co-deposits will be available to assist in the validation of the tritium monitor diagnostic.
However, an issue with analyzing the H content from surface analysis techniques is that themeasurements are complicated by the existence of extrinsic sources (such as atmospheric water
when the surface is exposed to air), which is why laboratory experiments use deuterium instead.
The use of trace deuterium during this experiment will therefore strongly increase the accuracy ofthe post-mortem analyses. A suitable L-mode target discharge has been identified (see Appendix A)
for such a migration experiment. Simulations indicate that the tritium production rate and in-vessel
activation will be low enough to allow a meaningful number of discharges to be executed.
To determine the feasibility of these experiments, DIVIMP and WALLDYN simulations have been
performed for the discharge described in Appendix A (Ip = 7.5 MA, Bt = 2.65 T (q95 = 3),
PIN = 30 MW, fd = 0.375, ne,sep = 11019 m-3). Two assumptions have been made to describe the farSOL and extend the plasma solution right up to all plasma-facing surfaces: either an exponential
extrapolation of the SOLPS solution (case 00d) or an increased radial velocity of 100 m.s-1 toaccount for the existence of a high density shoulder (case 00l). The RACLETTE code has also beenused to calculate the surface temperature profile in the divertor from the heat fluxes calculated fromSOLPS.
Figure 2.5-11 shows the Be deposition rate in the divertor for the low and high far-SOL density,
respectively. At the inner divertor baffle (corresponding to the area of interest for the T-monitor),
deposit growth rates in the range 0.01-0.5 nm.s-1 are predicted, the highest deposition rates being
found for the higher far-SOL density case. A measureable peak Be deposition ~1 μm would beobtained for plasma durations of the order of 104 s, or about 100 discharges with 100 s flattop.

Figure 2.5-11: Beryllium deposition rate in the ITER divertor calculated with DIVIMP/WALLDYN for an
L-mode plasma with Ip = 7.5 MA, Bt = 2.65 T (q95 = 3), PIN = 30 MW, fd = 0.375, ne,sep=11019 m-3, and a
low density far-SOL (left panel) and a high density far-SOL (right panel).

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These estimates have been made for the baseline (burning plasma equilibrium) equilibrium, but adecrease of the outer wall gap could be considered to increase the Be source and deposition rate into
the divertor region. The predicted retention rates are 210-4 gs-1 and 10-3 gs-1 respectively, so thatbetween 2 and 10 g of fuel would be retained in the vessel after 100 discharges much higher thanthe minimum measurable retention by the T-monitor, which is specified for a minimum in-vessel
retention of 1 g. Development time for this discharge would be required. Note that the discharge
flattop durations for the scenario explored in Appendix A are indeed on the order of 100 s. The useof trace D also significantly increases the accuracy of the gas balance analysis since background D
levels are extremely low. A full gas balance analysis can therefore be executed during thisexperiment. Cryopump regeneration will be performed before and after the discharge campaign,
which would be simply a repeat of identical discharges until the required accumulated discharge
time has been attained. The amount of D in the exhaust gas, after regeneration of the cryopumps,
can be very accurately measured using gas chromatography.
Importantly, this dedicated migration experiment can be used as a first validation of the fuelremoval strategy. After the experiment, ICWC conditioning can be applied to recover D from thewalls, and can be followed by a 350C divertor bake (walls at 240C). Using gas balance toestimate the in-vessel retention after each step (plasma discharges, ICWC, baking) would permit anestimation of the efficiency of the different techniques, whilst the post-mortem analyses would
validate the estimates of remaining D in Be co-deposits formed in the divertor.
In summary, the following methodology is proposed:







Execute a series of repeated 7.5 MA/2.65 T hydrogen L-mode discharges at the end of
PFPO-2 with trace D injection to accumulate divertor fluence and build measurable
deposition in the divertor;
Perform a gas balance analysis over the trace D discharge series to accurately estimate the Dretention rate;
Use the T-monitor to evaluate the thickness and fuel content of co-deposits on the innerdivertor baffle;
Apply ICWC discharges in hydrogen to recover D from the wall. Assess the amount ofrecovered D from gas balance analyses and local measurements using the T-monitor;
Perform a bake-out of the PFCs (240ºC for the FW, 350ºC for the divertor) and assess theamount of released D from gas analysis and measurements with the T-monitor;
Extract a divertor cassette during the following assembly phase for post-mortem analysis.

It is important to mention that the use of trace-D is also desirable for the TBM Program in order toevaluate the permeation rates of fuel species from the plasma to the TBM cooling systems (seesection 3.2.2). More detailed calculations and assessment of the experimental detection limits are
required to establish if the type of discharge proposed in Appendix A and the total discharge timerequired for the migration study would also be suitable for the TBM Program.
2.5.5.12.2.2 Ammonia formation during nitrogen-seeded H-modes
The higher available input power during PFPO-2 should allow the investigation of ammonia
formation rate during nitrogen seeded H-mode discharges. Since such discharges are likely to berequired for testing of divertor power load control, ammonia formation rates can probably be
studied without the need for dedicated discharges. Assuming that plasmas executed during thisphase will have higher divertor heat loads and higher wall fluxes, it should be possible toassess/confirm the role of these parameters on the ammonia formation rate. The approach used willbe identical as that described for the PFPO-1 (section 2.5.4.10.3.1). The main requirement will be to161ITR-18-003

perform a high temperature regeneration of the divertor cryopumps to recover the ammonia trappedon the active charcoal panels.
2.5.5.12.2.3 Dust generation and characterization
In common with the in-vessel tritium retention, there are inventory limits for the amount of dust
inside the vacuum vessel. The maximum permitted amount of mobilizable dust in the vessel is1000 kg. To prevent the risk of hydrogen explosion following the accidental ingress of air and water
into the vessel, the amount of dust on hot surfaces must be kept below 18 kg (9kg Be, 9kg W). It istherefore necessary to be able to diagnose the vacuum vessel dust inventory and to ensure thatsafety limits are not exceeded.
Several mechanisms may contribute to dust formation. Amongst them, the most significant are
expected to be droplet ejection during transient events (especially from melting of the Be wall) andflaking of layers forming as a result of first wall erosion, material migration and subsequentdeposition. Uncertainties remain as to how much dust can be generated during a disruption because
it is currently unclear how much of the melted material can be ejected in the form of droplets andthe proportion of these droplets that can convert into dust. Recent simulations have in fact shownthat droplets could be ablated in the disrupting plasma which would potentially strongly reduce thedust formation rate. Given that the wall heat loads during disruptions in ITER will be higher than in
any current device, it is unlikely that reliable, experimentally validated extrapolations will bepossible before the start of ITER operations. It is therefore important to quantify the dust production
rates in ITER, and to determine the characteristics of the produced dust (size distribution,
composition) prior to the start of nuclear operations.
An endoscope-based dust monitor system is being designed to view and collect particles below thedivertor (both on the cassette body and on the vacuum vessel floor). The system is planned to
become available during PFPO-2 and will allow regular inspection during the campaign. In parallel,
regular monitoring of the wall erosion/damage will be possible using the In-Vessel Viewing System(IVVS), which has the capability to perform direct viewing and measurements of height changes(with ~85% coverage of the FW) on surfaces, should be foreseen to identify the locations ofdamage formation and help adapt plasma scenarios to prevent further damage. The IVVS will also
allow the monitoring of excessive erosion, either due to direct sputtering/evaporation or due tomaterial redistribution from melt layer motion. A full vessel inspection requires about 8 hours andis therefore most likely to be performed during short-term maintenance periods.
Deployment of the dust monitor and IVVS should probably occur after a disruption or sequence ofdisruptions in which camera observations show particles being released. No dedicated experimental
time is required but the required deployment time must be factored in the planning. After PFPO-2, acomplete dust collection through vacuum cleaning of the divertor should be performed to quantify
the dust creation during the whole campaign. Dust collection on the vacuum vessel floor is notpossible unless divertor cassettes are removed so that if a cassette is removed for post-mortem
surface analysis (sections 2.5.5.12.2.1 and 2.5.5.12.2.4), this would provide an opportunity toinspect the vessel floor in this location and possibly aspirate debris before active operations begin.
2.5.5.12.2.4 Post-campaign analysis of Plasma-Facing Components
The divertor plasma-facing material in ITER will be exposed to unprecedented ion fluences andpossibly to a very large number of heavy transient heat loads. Damage (melting, cracking) could
occur early during operations, and it is important to document its occurrence and evolution. High
resolution inspection of the high heat flux regions of the divertor targets will provide information on
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the surface morphology and on the occurrence of melting and melt motion. The planned Erosion
Monitor (based on Speckle interferometry) can be deployed at the end of each operation day andprovides high resolution measurements of the surface morphology at a single toroidal location tocharacterize surface roughening (caused by melting or thermal shocking). This rather limited
coverage, however, calls for additional characterization.
A further concern with the tungsten divertor is to which extent early operations, including those inhelium, have modified the W thermo-mechanical properties and how this might impede later higher
power operations. To date, no such characterization is foreseen. Removal and full inspection of adivertor cassette after PFPO-2 appears to be the most straightforward option, and, as mentioned in
section 2.5.5.12.2.1, would greatly contribute to the characterization of Be migration. Once
removed, samples can be extracted (destructively) from the plasma-facing armour and sent out for
dedicated analyses. The results of this characterization would indicate whether the W
material/components are still compatible with the expected loads during high power operations. In
addition, as mentioned in section 2.2.10, removal of a cassette would allow a full characterization of
the erosion/deposition pattern in the divertor, in a way similar to what is performed at JET after
each campaign. Finally, as in PFPO-1, removal of the First Wall Samples will be performed after
PFPO-2 for analysis to determine erosion by charge exchange neutrals and retention by
implantation.
The exact strategy with respect to post-mortem analysis of samples (including dust) extracted fromthe machine is currently undecided. In 2014, a workshop was held at ITER Organisation (IO) toreview the suite of diagnostics foreseen for tritium retention/dust/erosion measurements. The panel
recommended the study of possible on-site facilities allowing the evaluation of erosion/deposition,
surface composition and fuel retention of first wall samples and dust [Philipps, 2014]. As a result, a
conceptual design has been developed for a laboratory to be located within the Hot Cell building[Galliez, 2017]. A formal decision on the actual implementation has not yet been taken. In parallel,
contacts have been made with several laboratories specialized in analyses of radioactive materials.
2.5.5.12.3 Heat loads during divertor operations
Experiments in the PFPO-2 phase will access for the first time heat loads approaching thosepossible during FPO when additional heating is supplemented by fusion power. L-modes inhydrogen with close to maximum PIN (73 MW), Bt = 5.3 T and high current (up to 15 MA) coulddeliver values of PSOL which can drive separatrix parallel heat fluxes to values up to ~50% of thoseto be expected in baseline burning plasmas, depending on the assumed values of SOL heat flux
width (see Heat and Nuclear Load Specifications [ITER_D_2LULDH, 2009]) and plasma density.
At these levels of q||, divertor heat flux control becomes mandatory depending on the achieved
divertor neutral pressures, though the lack of ELM activity means that the scenarios will not be
equivalent to a high power H-mode in terms of testing power load control in a fully integrated sense
for future operation in D and DT plasma.
As discussed in sections 2.5.4.9 and 2.5.5.7, there are various options for H-modes at lower Bt andIp, but the lack of an ICRF coupling scheme at 2.65 T in hydrogen probably means recourse to 1.8 T
and thus lower Ip or more exotic 3-ion absorption schemes at slightly higher Bt (3.0 – 3.3 T) forhydrogen H-modes. High power H-modes in He, which should be possible at 7.5MA/2.65 T, areattractive for first wall and divertor power load characterization, including the contributions of
ELMs, but much less attractive for divertor heat load control since the dynamics of divertor
detachment differ substantially to those in hydrogenic plasmas due to the very different divertor
radiation in He (section 2.2.1). In this sense, high power L-modes in hydrogen may be preferred forthe testing of impurity seeding for detachment control in this phase.
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In general, PFC heat load experiments in PFPO-2 should follow the methodology of those in
PFPO-1 (section 2.5.4.10.5), studying L-mode power widths, divertor spreading factors and general
detachment behaviour for a larger range of Ip at higher input power and Bt. Such studies benefit
from the addition of the divertor Thomson scattering in PFPO-2. Characterization of inter-ELM andELM divertor power loads and SOL widths can expand with the further development of hydrogen
H-modes at 1.8 T and in He and possibly in hydrogen at 7.5 MA/2.65 T. One of the key aims here isto establish a firmer scaling for the SOL heat flux width over a wider range of Ip, with H-mode apriority, but also recognizing the strong link between the L and inter-ELM H-mode power widths(section 2.2.2). Main chamber power loads should also be studied as in PFPO-1 over the wider
parameter range which will be available in PFPO-2, using scans in separatrix separation and plasma
density to study profile broadening and possible links to detachment.
Divertor power flux control methods, tested first in PFPO-1 should be further developed at thehigher power flux densities achievable in PFPO-2 L-modes at higher Ip and Bt. Seeding of both N2
and Ne can be tested, with the L-mode a more benign regime for such higher power experiments,
but in which SOL opacity will still require the use of core fuelling in conjunction with edge gas
injection of fuel and impurity. Such experiments must be conducted in hydrogen plasma to be of
relevance owing to the very different radiation distributions in He. Tests of the divertor plasma
response to loss of divertor seeding and asymmetric toroidal divertor injection will be required.
Assuming a robust hydrogen target discharge is available at higher power in PFPO-2, more
relevant, integrated detachment control of ELMy H-mode N2-seeded plasmas can be further
developed from the initial work in PFPO-2, but will again be limited in the range of density which
can be studied if these plasmas have to be operated at low field and current. Having a full set of
ELM control coils power supplies in this phase should allow optimization of the 3-D spectrum andtests of rotation of the perturbation so that further measurements of divertor heat load patterns canbe performed, even if the density will be limited for low-Ip H-modes.
2.5.6

Summary of experimental activities and operational time in the PFPO Phase

An ambitious experimental program is foreseen for the PFPO phase, advancing the facility
operational capabilities from sustainment of the simplest plasma breakdown, with plasma currentsof less than 1 MA, to fully integrated L- and H-mode plasma scenarios in the range 5 – 15 MA
which exploit essentially all of the tokamak, auxiliary and plant systems. The overall objectives ofthis phase are to demonstrate essentially the full technical capability of the ITER systems (acceptingthat the limiting capability in areas such as heat loads, pulse length etc., might not be accessible
until significant fusion power is produced), to achieve robust and routine plasma operation over theaccessible L- and H-mode parameter range, to perform initial exploration of fusion plasma physics
at the ITER scale, and to prepare the scientific, technical and operational basis for the transition to
DT operation, leading to significant fusion power production.
The PFPO research program will already allow extensive testing of many elements of the physics
basis developed to support the ITER design and performance predictions. The values of key physics
parameters such as collisionality and normalized Larmor radius, as well as ‘engineering parameters’such as plasma current, major/minor radius etc., accessible during this phase will extend thetokamak physics basis very considerably and allow numerous comparative studies of predictions
derived from the ITER physics basis in plasma transport and confinement, MHD stability, divertor
and plasma wall interactions, heating and current drive efficiency, plasma control and scenario
integration, and possibly also energetic particle physics (but still with anisotropic velocity
distributions). Perhaps the key caveat in this respect is that, although a substantial program
addressing H-mode physics (access conditions, energy and particle confinement, ELM
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characterization and control etc.) is planned, the extent to which key aspects of H-mode physics canbe addressed adequately will probably depend on achieving satisfactory H-mode performance inhydrogen as well as helium – key physics tests in this area might otherwise await the transition todeuterium operation in the FPO phase. Overall, therefore, and despite the inevitable emphasis ‘oncommissioning with plasma activities’ within the presentation of the PFPO research program, thisphase of ITER operation will provide significant opportunities for studies of fusion plasma physics
at the ITER scale and will allow access to plasma and operational parameters substantially beyond
those of current experiments.
The experimental program implemented during PFPO will also provide wide-ranging tests of muchof the ITER technology under challenging conditions. It will, in particular, launch the first phase of
the TBM testing program (see section 3.2.2.2), a central element of the ITER mission. Operation ofmajor systems and components such as the superconducting magnets, plasma facing components,
heating and current drive systems, Diagnostics, fuelling and vacuum systems, control systems,
power supplies, the cryogenic system, tokamak and auxiliary cooling water systems (and evenremote handling systems, which are required in this phase simply due to the scale of ITER
components) will already yield an invaluable database on system performance and reliability overpulse lengths ranging from several tens to several hundreds of seconds. The upper limit of thisrange in pulse length will rely on successful commissioning and operation of the tokamak, auxiliary
and plant systems, as well as successful implementation of the scientific research program, which
will determine the rate of progress through the various performance milestones, potentiallyreleasing time for commissioning of the key tokamak and facility systems, in particular, the H&CD
systems, to extended pulse lengths of several hundred seconds.
The routine operation of plasma pulses with long durations during the PFPO phase would be adesirable, and significant, achievement, which would give additional confidence in operationalreliability in advance of the FPO phase. This would be a major advantage in the preparation for thetransition to DT operation, and could be expected to shorten the experimental time required toachieve long-pulse operation at high fusion power in DT plasmas. Nevertheless, if all otherelements were in place to allow a timely transition to D/DT operation in late 2035, as scheduled,reliable L- and H-mode operation over a range of plasma scenarios with flat-top durations of several
tens of seconds would be an adequate criterion to support this transition, given the emphasis on
ITER’s role in demonstrating fusion power production as rapidly as possible.
Table 2-9 and Table 2-10 provide an overview of the experimental activities implemented in eachof the PFPO phases and of the operational time estimated as necessary to meet the objectives
defined for each phase. There are, of course, significant uncertainties in these estimates, given thecomplexity of some of the ‘commissioning with plasma’ activities which it is planned to implement
during the PFPO phase and the reliability issues which might be encountered in commissioning
plant and auxiliary systems for the first time. For comparison, the estimated experimental timerequired for PFPO-1 is between 412 and 468 days (including the programmatic options considered)compared to an estimate of 470 operational days available (see Table 2-1), implying that there islittle contingency in the estimated time schedule. For PFPO-2, the proposed program is estimated torequire 418 – 433 days, as compared to the estimated 545 operational days available. Thus the
PFPO-2 program does include contingency against unexpected developments in the program. Thispotentially offers opportunities to place a greater emphasis on the commissioning of H&CD
systems to long-pulse operation and to exploit this capability in the experimental program.
Bringing all ITER systems from the initial phase of Integrated Commissioning to preparedness for
DT operation, establishing the required range of plasma scenarios and demonstrating the required
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Table 2-9 – Overview of experimental activities and estimates of required operational days for
the PFPO-1 research program
Activity
Days
Plasma control/ Plasma scenario development
155
PF Null/ Breakdown optimization
7
Commissioning of gas and pellets for density/impurity control
9
Scenario development/ plasma magnetic control to 3.5 MA Limiter
17
Scenario development/ plasma magnetic control at ~5 MA/2.65 T Divertor
32
40
Scenario development/ plasma magnetic control at ~7.5 MA/2.65 T Divertor
50
Scenario development/ plasma magnetic control at ~7.5 MA/5.3 T Divertor
Disruption management
64
Disruption load validation
10
Commissioning/ Optimization of disruption prediction/ detection
11
Commissioning/ Optimization of disruption mitigation
43
Commissioning of auxiliary systems
75
Diagnostics
10
ECRH commissioning at 170 GHz (plus ~11 days for First Plasma)
25 [75]*
ECRH commissioning at 104/110 GHz
10 [30]*ICRF commissioning
30 [60]*Commissioning of advanced control functions
79
Advanced control functions (wall/divertor protection, MHD control etc.)
54
Termination scenarios, exception handling, shared actuator management
25
Divertor and PWI studies
39
Divertor/wall heat loads in divertor configuration and detachment control
5
Limiter heat loads
15
Disruption heat loads
5
Wall cleaning
10
PWI studies (fuel retention, He effects etc.)
9
Option: 1.8 T H-mode studies
44
H/He H-mode access at 5 MA/1.8 T (including ELM control)
31
Studies related to TF ripple effects on H-modes (preparation for TBM effects)
10
Assessment of H-mode power threshold variation with Bt
3
Option: Extension of plasma operation towards 10 MA
12Total
412 [+56]* The total estimated time for initial commissioning is shown in brackets, with an allocation of 1/3 to 1/2 of the time to
dedicated experiments.

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Table 2-10 – Overview of experimental activities and estimates of required operational days
for the PFPO-2 research program
Activity
Axisymmetric magnetic control
Plasma scenario commissioning
Kinetic control commissioning
MHD and error field control
Advanced control functions (auxiliary specific, actuator sharing etc.)
Disruption Management
Disruption load validation
Commissioning/ Optimization of disruption prediction/ detection
Commissioning/ Optimization of disruption mitigation
Commissioning of auxiliary systems
Diagnostics
ECRH commissioning
HNB + DNB commissioning
ICRF commissioning
Divertor and PWI studies
Divertor/wall heat loads in divertor configuration and detachment control
Limiter heat loads
Disruption heat loads
Material migration
Fuel retention/ management
PWI studies (wall cleaning, He effects etc.)
H-mode studies
H-mode studies at 5 MA/1.8 T (TBM effects)
Expansion of H-mode operation towards 7.5 MA/2.65 T, incl. ELM control
Development of long-pulse H-modes
Development of 15 MA/5.3 T L-mode
Current drive efficiency, target q-profile formation, fast particle effects
Total

Days
5
10
50
30
34
27
1
6
20
70
10
10
25 [50]*25 [50]*
40
7
15
2.5
5
5.5
5
111 - 12640
65 - 80625
16
418 - 433

* The total estimated time for initial commissioning is shown in brackets, with an allocation of fraction of the time to
dedicated experiments.

level of operational reliability to support the transition to the FPO phase within the scheduled periodof experimental operation of ~40 months is, as noted above, ambitious. There is a range of risks, interms of hardware reliability/performance and plasma performance (as discussed in preceding
sections and consolidated in Appendix J), which, if encountered, will need to be addressed andresolved, or mitigated, rapidly to ensure that the schedule towards DT operation proposed within
the Staged Approach is maintained. Many of these risks are being addressed in ongoing physics andtechnology R&D programs. Nevertheless, given the scale of the extrapolation from current devices
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to ITER, the associated uncertainties in the underlying physics basis, and the novelty of elements of
ITER’s integrated plasma scenarios, it is inevitable that expectations of plasma behaviour andperformance, in particular, must be tested in the ITER environment. As indicated in the discussion
in previous sections, a significant outcome of these tests and of measures to address theconsequences of risks realized in the research program will be the specification of system upgrades
required to improve operational reliability and assure the required plasma performance in later
phases of the Research Plan (see also the detailed discussion of potential upgrades developed in
section 4).

2.6 Fusion Power Operation Phase (FPO)
The Fusion Power Operation Phase, as discussed in this version of the Research Plan, encompasses
the first experimental campaign of the FPO phase, which involves both D and DT plasmas andfocusses on the development of inductive plasma operation to fusion power production levels of
several hundred megawatts, and the second and third campaigns, in which long-pulse operation isdeveloped in inductive, hybrid and non-inductive DT plasma scenarios, with the aim of meeting the
project’s mission goals of Q ≥ 10 for periods of 300 – 500 s, and Q ≥ 5 in fully non-inductive
operation for periods of up to 3000 s.
2.6.1

Objectives for the Fusion Power Operation Phase

During FPO, ITER will operate with all its baseline ancillary systems at full performance, including
the T-Plant. In this phase, the three ITER DT high-Q goals should be successfully demonstrated.
The initial focus of the research will be towards the achievement of Q ≥ 10 for ~ 50 s, following
which it will shift to extending the Q = 10 burn regime towards the 300 – 500 s goal andsubsequently to developing the long-pulse and steady-state scenarios towards Q = 5. It is expectedthat upgrades to the ITER ancillaries will be available for operation, either from the start of FPO(e.g. HNB3) or soon after the initial years, if the upgrades are required to achieve the Q = 5 goal,
but are not implicated in achieving Q = 10 (e.g. additional ECRH power, RWM control power
supplies, etc.). Similarly the T-Plant will be commissioned and will start providing significant
quantities of T for the experimental program ~ 1 year after the start of FPO, allowing an extended
program of DT operation to start.
The FPO phase can be divided into four blocks.
Initial D and trace-T operation:
In this phase all ITER systems will be commissioned (or recommissioned) for nuclear operation andthe target scenarios of PFPO-2 will be reproduced in D plasmas. This implies the recommissioningof the ITER H&CD for D operation (mainly the HNBs, for which the injected species will change
to D, and ICRF), diagnostics (chiefly the neutron-related diagnostics), fuelling (change to D), etc.
In this phase the plasma conditions achieved in H and/or He plasmas will be reproduced, namely
L-mode operation will be expanded to 15 MA/5.3T and H-mode operation will be explored up to
7.5 MA/2.65 T. The operational strategies and control schemes developed during PFPO will beapplied and refined/ retuned to the higher plasma energies expected in D plasmas. This phase will
provide first experimental confirmation of the plasma parameters expected in DT plasmas in thiscurrent range, as the plasma characteristics are expected to be similar in D and DT plasmas except
for those related to high fusion power production. Once D operation is reliably established, trace-T
experiments will be performed with special focus on the 7.5 MA/2.65 T H-mode scenario, including

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also the L-mode phases. Tritium transport and fuelling experiments will be performed and theresults will be used to optimize fuelling and mixture control strategies for DT.
In this phase, research to confirm the key ingredients of the long-pulse/ steady-state scenarios will
be confirmed. This will include experiments similar to those in PFPO-2 to develop the target-q
profiles in the ramp-up, to sustain them for ~ 10 s in the flat-top and to determine the current drive
efficiency of each of the H&CD systems (chiefly NBI and ECRH) in H-mode plasmas. The extent
to which these experiments will be performed will be determined by progress on the H-mode
program and the availability of the T-plant for DT plasma operation.
The main objectives of this operational block are:













Commission plasma control, interlock and safety systems with plasma to the required levelto support the experimental program;Commission the additional diagnostic systems (particularly those related to neutrons),
validation of data and integration into control/ interlock systems;
Recommission the installed H&CD systems for D plasmas to their installed power level forat least 50 s;
Extend the capability for disruption prediction, avoidance and mitigation to parameterscharacteristic of the highest performance plasma achievable in D plasmas;
Demonstrate D L-mode operation to 15 MA/5.3 T;
Demonstrate robust NTM control in high confinement D H-mode plasmas;
Develop D H-modes to 7.5 MA/2.65 T (~50 s) with reliable ELM control and stationary
divertor power load control;
Physics characterization of ohmic, L-mode and H-mode plasmas in D;
Perform trace-T experiments in D L- and H-mode plasmas to assess T transport and T
fuelling requirements and schemes;
Characterization of key aspects of plasma-wall interactions in the ITER environment in D
plasmas;
Preparatory studies of hybrid/ non-inductive scenarios in D plasmas;
Preparatory studies of ‘burning plasma’ issues (e.g. helium exhaust, fuel mixture control in
trace T, etc.).

Initial DT operation with non-trace T level:
In this phase, the amount of T provided by the T-plant should allow operation with T concentrationsin the range of 10-70% for up to ~ 50 s. Fusion energy production will start to be significant andneutron measurements will start to be important to evaluate it, as well as for the execution of theexperimental program. As a consequence of the fusion energy production, the effects of neutron
heating on the superconducting coils will start to be measurable and, although this is not expected toraise any issues for operational durations of ~ 50 s, the consequences for longer pulse operation athigh Q will need to be evaluated at this stage. This may call for further optimization of the scenarios
to be implemented at this stage to reduce AC losses. With the increased level of power in DT Hmodes, routine operation of the scenarios with ELM control and radiative divertors will be required;this is expected to necessitate development and/or retuning of the control schemes developed for D
at lower currents, with the same applying for the disruption avoidance and mitigation schemes.
Similarly, the T throughput will start to be non-negligible, and already in this phase T-retention
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measurements and removal schemes are expected to become part of the operation cycle inpreparation for their routine use in the next operational DT phase.
The plasma regimes to be explored in this phase start with 7.5 MA/2.65 T plasmas, which will becharacterized extensively both in L-mode and H-mode over a range of T concentrations. In theseexperiments, the first effect of -heating will be documented, as well as fuelling requirements andpossible new edge-core integration requirements that could arise due to the present of T in theplasma. Once a fully integrated 7.5 MA/2.65 T H-mode scenario in DT has been developed, theplasma current/toroidal field will be increased in steps towards 15 MA/5.3 T, and the demonstrationof Q = 10 for ~50 s. It is expected that the steps will be similar to those in the L-mode D program,
except that additional steps in current/ field will be added between 12 and 15 MA to account for thelarge change of plasma energy with Ip in H-mode. For each step, the T concentration will bescanned and fully integrated scenarios will be developed, including ELM control, fuelling anddivertor power load control. It should be noted that the range of T concentrations that will be
explored will decrease as the plasma current increases because the -heating will become an
essential part of the scenario to sustain the high confinement H-mode. It is expected that, in the
expansion towards high-Ip/ high-Q, control of MHD stability will become more challenging andsignificant time will have to be dedicated to the control of NTMs and sawteeth in this phase.
In this phase, research to perform an initial development of Demo-relevant long pulse scenarios will
start. This will include current ramp optimization in DT L-modes and H-modes (during the rampup), as well as sustainment of the flat-top H-mode phase. It is likely that, in this first operational
phase, the sustainment will be limited to a few tens of seconds with relatively low Q, due to thelimited operational time available and the available H&CD schemes.
The main objectives of this operational block are:












Commission plasma control, interlock and safety systems with plasma to the required levelto support the DT experimental program towards Q = 10;
Commission additional diagnostic systems (e.g. for fusion power and fusion products),
validation of data and integration into control/ interlock systems, as required;
Commission ICRF heating scenarios for DT plasmas to installed power level for at least
50 s;
Extend capability for disruption prediction, avoidance and mitigation to parameterscharacteristic of the highest performance in DT plasmas;
Develop DT H-modes at 7.5 MA/2.65 T (~50 s) with reliable ELM control, NTM control
and radiative divertor operation;
Expand DT H-mode operational space from 7.5 MA/2.65 T plasmas culminating in
12- 15 MA/5.3 T DT H-mode operation with full feedback control at fusion powers of
200 – 300 MW for at least 50 s;
Extend operation to 15 MA (if not done above) and optimize fusion gain to demonstrate
Q ≥ 10 for at least 50 s;
Optimize fuel mixture control, burn control and plasma-facing component protection in DT
plasmas;Physics characterization of ohmic, L-mode and H-mode plasmas in DT;
Explore options for current profile control in DT plasmas to identify most promising
approaches for long-pulse/ steady-state operation.

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It should be noted that it is not expected that the initial target of Q = 10 for ~50 s will be achieved in
the first FPO campaign: the foreseen target for the end of the first FPO campaign is a fusion power
target of 200 - 300 MW with a fusion gain of Q ~ 5 sustained for ~ 50 s. Expanding this firstly
towards a scenario with Q ≥ 10 for 50 s and then extending the burn duration to 300 - 500 s is
expected to occur after the first experimental FPO campaign lasting 16 months.
Second phase of DT operation:
In this phase, the Q = 10 scenario will be extended towards 300-500s and the long-pulse and steadystate scenarios will be developed. The rate of progress of these developments will depend, ofcourse, on the results of the experiments and on the physics itself, as well as on the status of
implementation of upgrades which may be required to achieve the Q ~ 5 goals (e.g. additional
ECRH/ECCD, power supplies for RWM control, etc.). During this phase the T throughput willreach its maximum value and T retention measurement and removal schemes will be routinely
applied as part of the operation cycle. Similarly, neutron heating of the TF coils is also likely toreach its maximum, and this may require re-optimization of the scenarios to decrease AC losses.
With the increased burn length, fusion power/ power load to components and inductive current
fraction, plasma control in general and event handling, in particular, are expected to become
significantly more complex and a substantial fraction of the operational time to extend the burnlength is expected to be dedicated to solving such issues.
The main objectives of this operational block are:
















Develop DT plasma scenarios with Q ≥ 10 at Pfus ~ 500 MW for 300 – 500 s;
Characterize key aspects of PWI in the ITER environment in high fusion power/ high gain
plasmas;Develop program of burning plasma studies around high fusion power/ high gain inductive
plasmas;Demonstrate satisfactory tritium accounting, retention measurement and removal
techniques;
Provide required support to the TBM testing program;
Investigate optimization of fusion power with Q > 10 during stationary inductive conditions
and characterize plasma behaviour in these conditions;
Identify needs for upgrades to improve fusion performance, or to improve potential for noninductive operation (e.g. ELM control coils power supplies for RWM control, if notpreviously possible to do so);Establish robust current profile control techniques for DT plasmas;
Establish DT plasma scenarios with current profile control, improved H-mode confinement
having a burn duration of ~ 1000 s and several hundred megawatts of fusion power basedon the hybrid/improved H-mode scenario at q95 ~4;Establish non-inductive plasmas with the longest sustainable pulse duration (> 1000 s) andmoderate fusion performance (Q ~ 2);
Optimize fusion gain and pulse duration to establish plasmas with durations beyond 1000 s
and Q ≥ 5;
Optimize Q in fully non-inductive DT plasmas towards Q ~ 5;Start the characterization of key aspects of PWI in the ITER environment over very long
durations (e.g. impact on fuelling, erosion/ redeposition, fuel retention, etc.).
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It should be noted that, depending on the suitability of the upgrades implemented by this phase andthe need for additional measures to achieve the non-inductive goal, the latter part of the researchand development in this phase may be implemented in the next and final phase of FPO.
Long-term DT fusion R&D research:
In this phase, R&D issues related to the long-term goals of ITER towards Demo will be addressed.
This is likely to require specific Demo-related upgrades discussed in section 4, which will beidentified and implemented in this phase, together with the completion of those required to achieve
the Q ~ 5 non-inductive goal.
The main objectives of this operational block are:




2.6.2

Conduct a program of burning plasma studies to characterize physics of non-inductive DT
scenarios;
Perform systematic characterization of key aspects of PWI in the ITER environment oververy long durations (e.g. impact on fuelling, erosion/redeposition, fuel retention etc.);
Provide required support to the TBM testing program;
Identify options for further upgrades to prepare longer-term Demo-targeted studies.
Assumptions

The research program in FPO is based on the assumption that all the ITER baseline systems will befully operational (including H&CD, fuelling, etc.) with all ECRH gyrotrons at a frequency of 170
GHz. Table 2-11 lists the final elements of the baseline diagnostic systems which should beinstalled to satisfy the essential measurement requirements for the FPO phase. It also shows a rangeof upgrades which would enhance the studies of burning plasma physics and which could beinstalled if the necessary funding were available sufficiently in advance of the first FPO campaign.
It is furthermore assumed that the T-Plant will start to provide a significant throughput for DT
operation (at least 10% concentration for ~50 s) not later than one year after the start of D
operation. It would also be of significant benefit to the experimental program if all upgrades
identified in PFPO as required for achieving the Q = 10 goal were to be installed before the start of
FPO operation.
Regarding plasma scenarios, it is assumed that fully integrated scenarios have been developed for
L-mode plasmas up to 15 MA/5.3 T and for H-modes up to 7.5 MA/2.65 T (including ELM control,
power load control, etc.), which will require retuning, but not full redevelopment in D/DT. If thelatter is not achieved in PFPO due to difficulties in accessing the H-mode regime in H/He plasmas,
additional experimental time in D plasmas will be required to determine H-mode access conditions,assess H-mode plasma behaviour at the ITER scale and develop control schemes, as part of theintegrated scenarios assumed for D/DT. Similarly, it is assumed that disruption mitigation andavoidance schemes have been developed in PFPO for the same range of scenarios and that retuning,
but not full re-development, will be required for D/DT.
As discussed above, it would be desirable that upgrades identified as required to achieve the longpulse/ steady-state goals were already implemented before the start of FPO (e.g. additional
ECRH/ECCD power). Obviously this does not concern the possible upgrade of a 3rd HNB, forwhich certain measures must already be taken well before the start of FPO, in particular the
installation of the captive components, and whose installation, if required, must take place in thelast assembly phase (Assembly IV) before the start of FPO (see Appendix G). However, as the first
operational period will be focussed towards the achievement of the Q = 10 goal, and this is likely to172

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take more than the first 2 years, a later implementation of these upgrades (if not delayed by morethan ~3 years after the start of FPO) are not likely to severely affect the development of the research
program to achieve the Q ~ 5 goal.
Table 2-11 – Overview of additional measurement requirements and additional diagnostic
systems installed for the FPO-1 campaign*
Measurement Requirement

Installed Diagnostic Systems• Radial Neutron Camera
14 MeV neutron diagnostics and neutron profile• Vertical Neutron Camera
cameras• Neutron Calibration (14 MeV)
Confined- and lost- diagnostics**• Collective Thomson Scattering• Lost Alpha Monitor

Additional fusion product diagnostics**

•
•
•
•

PWI/ Heat Load characterization• Tritium Monitor (fully operational)

High Resolution Neutron Spectrometer
Tangential Neutron Spectrometer
Vertical Gamma Ray Spectrometer
Radial Gamma Ray Spectrometer*

The table illustrates the additional diagnostic systems planned for installation prior to FPO to meet the additional
measurement requirements. The installation planned for Assembly Phase IV should essentially complete the Diagnostic
baseline specification. However, a variety of upgrade options to extend the baseline measurement capability has been
identified for which funding might become available on this timescale – see Appendix H for a discussion of these
upgrade options and Section 4 for a summary of the possible benefits to the research program.
**All diagnostic systems in these categories are currently categorized as unfunded upgrades.

2.6.3

Operation plan for deuterium plasma experiments

2.6.3.1 Plasma restart in FPO, including H&CD and Diagnostic commissioning
The plasma restart in FPO should establish routine plasma operation in deuterium at 7.5 MA/2.65T
after the installation and first commissioning of the tritium plant and completion of the final
preparations for DT operation. This is the first time that diagnostics will see significant neutron
fluxes; therefore, careful commissioning of the shape and current control, vertical position control,
and first wall heat flux management systems must be carried out at the beginning of FPO. The main
restart activity for the H&CD systems will be the commissioning of the NB systems for deuterium
operation.
Plasma initiation in deuterium will be slightly different from hydrogen and helium operation in theprevious phases. Once a robust initiation is found at 2.65 T and 5.3 T, the only further dedicatedtime to work on plasma initiation that may be required is when the full CS performance is required.
The optimized initiation sequence at reduced CS capability may make use of coil current
combinations that are not allowed at maximum precharge of the CS due to force or field limits.
Optimization of the plasma initiation can be done parasitically during other restart activities.
Recommissioning of the plasma shape, current and vertical position control will follow theprocedures developed in the previous phases. The key issue is to look for effects of radiationinduced voltages in the sensors. Precise shape control will be needed for recommissioning of thefirst wall heat flux protection similar to the protocols developed in the previous phases. The
milestone for this element of the restart is robust 7.5MA/2.65 T operation.

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Density control will need to be retuned to account for the different transport behaviour of deuterium
neutrals and ions from previous experience in hydrogen and helium. Integrated use of the tritium
plant for supplying the gas and pellet injectors is necessary. Fuel mix control will have beendeveloped for ICRF and H-mode operation using H/He mixtures in previous phases. This will need
to be adapted for DT mix control, plus minority concentration control for ICRF.
The DMS system should be recommissioned using reference pulses from PFPO-2 and any changes
to DMS hardware installed during Assembly Phase IV must be commissioned. Operation at 5.3 T
will be necessary to demonstrate that runaway mitigation is maintained in deuterium.
The major activity of restart of the H&CD systems will be the commissioning of the HNB systems
in deuterium. This will be the first operation at 1 MeV acceleration voltage and in the presence ofsignificant neutron flux. The change in working gas in both the beamline and the plasma will
require requalification of the internal protection systems in the beamline and port. The ICRF
coupling control will need to be re-optimized for deuterium. There should be no changes requiredfor ECRH operation. All three systems should be recommissioned to at least 50 s operation in orderto support the research program. If the systems were commissioned to longer pulse in PFPO-2, it
would be good to demonstrate that longer pulse capability again in the restart phase, if time permits.
Once H&CD power in excess of 40 MW is available, the divertor heat flux management system
will need to be recommissioned. The controls will need to be re-optimized for the different
behaviour of deuterium neutral and ion transport.
New control issues associated with the introduction of significant self-heating such as entry to burn,exit from burn, and robust normal and off-normal termination schemes will be developed within the
research phase.
2.6.3.2 Disruption management program in FPO
The disruption program in FPO will focus on monitoring disruption budget consumption and taking
corrective actions when needed. This also involves monitoring the performance of the DMS at
higher energies and with possibly additional RE seed mechanisms (Tritium decay, Compton
scattering). Priorities will be given to reducing disruption rates by appropriate event handling of the
PCS (see section 2.5.4.4) and by active MHD control (see section 2.6.3.3). The thermal quench
mitigation rate in FPO will be determined by the thermal loads themselves, but also by the need tosuppress runaway formation that could potentially follow from strong tungsten influx. The
requirements for the current quench mitigation rate will already be identified from the EM load andcurrent quench heat load scaling carried out in PFPO-2 and is expected to be close to 100%. Thiseffective current quench mitigation will be based on the disruption detection that has been
established in PFPO-2. Especially because of the thermal energies achieved in FPO, there might be
the need to further optimize the mitigation sequence or species composition to achieve higher
radiation rates, while keeping radiation peaking and thus heat loads to the first wall (FW) at
acceptable values. Note that the impurity quantity is limited by the need to keep eddy current loads
on blanket modules in the acceptable range during the current quench and it may not be sufficientfor thermal load mitigation at higher thermal energies. This may put further constraints on therequired disruption rate. On the other hand, it has been observed in experiments that the energy at
the time of the thermal quench is in many cases significantly lower than the maximum energy that
was achieved during that pulse. In PFPO-2 high energy thermal quenches are expected mainly for
VDEs. In FPO, depending on the plasma scenarios, other disruption types may become relevant thathave higher energies at the time of the thermal quench, also demanding higher TQ prediction rates.
Improving plasma performance can also introduce new stability limits (e.g. pressure driven
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instabilities). This, together with possible changes in diagnostic capabilities will require adapting
the disruption prediction schemes to ensure long enough warning times to activate avoidance
capabilities or the DMS.
2.6.3.3 Advanced control commissioning in FPO
Advanced control commissioning for the FPO phase will extend the advanced control
commissioning from the previous operational phases to high performance operation in D and DT
plasmas. It is not yet clear whether there will be new or upgraded actuators in FPO, but there will be
new diagnostics for fusion products, and the additional fusion power will enhance plasma
performance, requiring significant changes to advanced control schemes, particularly for kinetic
control, MHD control, integrated control, and event handling. Neutron and gamma ray emission
from fusion reactions may also impact some actuators and measurements so that additional
commissioning of advanced control functions that depend on the systems impacted may berequired. Note that considerations for disruption-related control in FPO have been already describedin section 2.6.3.2. The run-time estimates given here are for commissioning of advanced control
schemes, but further development of the physics R&D needed to develop robust control is likely torequire significantly more run-time as part of the physics program.
2.6.3.3.1 Plasma kinetic control
The additional load on the first wall due to fusion power in the D and DT phases up to several
hundred MW will require significant changes to first wall protection. With longer pulse durations
and higher fusion performance, first wall protection will become more challenging, requiring wellcommissioned robust control schemes. Additional commissioning of first wall protection has been
allocated 5 run days in the FPO phase.
Divertor protection with up to 73 MW from all three heating systems and several hundred MW offusion power will also be significantly more challenging in the FPO phase. At high plasma
performance, divertor heat flux control is essential for investment protection requiring wellcommissioned robust control schemes. Additional commissioning of divertor protection is allocated
5 run days in the FPO phase.
The plasma  and stored energy control will require changes due to the additional fusion power in
the FPO phase and it will be coupled to the fusion burn control scheme. Model-based control will
require the addition of -particle heating to the controller model. Additional commissioning of and stored energy control is allocated 2 run days in the FPO phase.
The temperature and pressure profile control system should be largely commissioned by the start of
the FPO period, but the additional fusion power will be concentrated in the plasma core and will
dominate the control scheme at high fusion power. It will also be coupled to fusion burn control.
Model-based control will require the addition of -particle heating to the controller model.
Additional commissioning of temperature and pressure profile control is allocated 2 run days in the
FPO phase.
The rotation profile control system should be nearly fully commissioned by the start of the FPO
period. There may be some impact of neutron and gamma ray emission on rotation profile
measurements that could require additional commissioning. Alteration of NTV due to possible nonambipolar loss of -particles or changes of the uncorrected error field due to the new TBM
implementation in the FPO phase may require changes to a model-based control algorithm.
Additional commissioning of rotation profile control is allocated 2 run days in the FPO phase.

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The current density profile control system should be nearly fully commissioned by the start of the
FPO period. There may be some impact of neutron and gamma ray emission on current density
profile measurements that could require additional commissioning. Increased plasma performance
will affect the central current density and the edge bootstrap current and affect current density
profile control. Additional commissioning of current density profile control is allocated 2 run daysin the FPO phase.
The FPO phase will finally permit the development of fusion burn control, which will use all
H&CD systems and DT fuel mixture control with gas and pellet injection as actuators. Impurity
injection may also be used to radiate energy and also control the fusion burn. Indeed, many of theother advanced control schemes will affect the fusion burn, so it will be a strongly coupled systemthat can only be studied once ITER reaches fusion burn conditions and will require model-based
control schemes. The control of the accumulation of helium ash from the fusion burn in the core
will also affect burn control, particularly for longer burn durations. Additional commissioning of
fusion burn control is allocated 2 run days in the FPO phase, but it will require much moredevelopment within the physics program to develop robust long-pulse burn control.
2.6.3.3.2 MHD and error field control
The ELM control schemes commissioned during the previous operational phases should still apply
to the FPO phase, though ELM control will be essential at high fusion performance and thus it must
be well commissioned and robust before high fusion performance is achieved. No dedicatedcommissioning time has been allocated to ELM control in the FPO phase, though the developmentof robust ELM control will be a significant part of the physics program.
NTM control schemes should already be developed to an advanced state by the start of the FPO
phase and should remain applicable. The additional plasma performance in this phase will drive
large sawteeth and ELMs that provide seed islands for exciting NTMs, particularly at high-.
Before high fusion performance can be achieved, effective, robust NTM control must be
commissioned to the extent that it becomes routine. An upgrade of the ITER ECRH system may beneeded prior to the FPO phase to ensure effective NTM control. To develop NTM control to fullmaturity, ITER will have to allow for substantial experimental time within the physics program.
The DEFC and RWM control system should be largely commissioned by the start of the FPO
phase. The new TBMs in this phase may change the intrinsic and dynamic error fields of the device
that may necessitate further commissioning to ensure that adequate correction is available.
Moreover, the use of deuterium and, with the transition to DT operation, the addition of significant
-heating will increase the plasma performance and raise the stored energy closer to plasma
-limits, leading to greater plasma amplification of error fields and potential RWM destabilization.
A model-based control system should be upgraded to include the changes to the TBMs.
Sawtooth control should be well developed by the start of the FPO phase. The additional plasma
performance and energetic -particles in this phase will drive large sawteeth and affect theircontrol, which is likely to be essential for maintaining high performance operation. The schemes
developed in the PFPO-2 phase should apply to some extent but, because energetic -particles arelikely to affect sawtooth stability, this will require significant changes to the control schemes. The
development of high fusion performance will include robust sawtooth control as part of the physics
program.
Initial Alfvén Eigenmode control will have been developed during the PFPO-2 phase, but theadditional fast -particles in the FPO phase will change significantly both the drive and damping of176

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Alfvén Eigenmodes. Significantly more R&D along the lines of [Van Zeeland, 2016] is needed on
existing devices to develop Alfvén Eigenmode control for ITER. The large -particle drive on
ITER, however, will probably require significant development of Alfvén Eigenmode control on
ITER itself through an extensive physics program.
2.6.3.3.3 Supervisory control
The supervisory control functions of the PCS should be well developed by the start of the FPO
phase. The additional plasma performance in this phase will probably lead to more significant
plasma events that will require additional commissioning to handle the required control changes andthe nuclear heating of some components such as the superconducting magnets will require
additional model-based exception handling to avoid exceeding operating limits. At high plasma
performance, exception handling becomes particularly critical for investment protection and needs
to be well commissioned and robust before high fusion power levels are reached. In particular, thethermohydraulic models for operating the superconducting coils with eddy currents combined with
nuclear heating must be optimized to ensure proper event handling with model-based control,
particularly to maintain high plasma performance for long pulse operation.
Shared actuator management will be largely developed prior to the FPO phase, but the addition offusion power and burn control will even more strongly couple all of the advanced control functions
that affect plasma performance. Thus the physics program will have to include further developmentof actuator sharing and model-based control as part of the high fusion performance development
program.
2.6.3.3.4 Deliverables for advanced control commissioning in FPO
The deliverables for advanced control commissioning in the FPO phase include:










Routine operation of first wall and divertor heat flux control with fusion power;
Implementation of thermohydraulic model-based control of the superconducting coils witheddy current and nuclear heat loads and event handling;
Establish fusion burn control, β, and stored energy control;
Commission temperature, pressure, rotation, and current density profile control;
Reliable ELM control with additional fusion power;
Sawtooth control with fast -particles;
Robust and routine NTM control with additional fusion power;
Alfvén Eigenmode control with fast -particles;
Commission dynamic error field correction and RWM control at high plasma performance.

2.6.3.4 Development of deuterium L-mode operation to 15 MA/5.3 T and first assessment of
H-mode access at 2.65 T in D plasmas
2.6.3.4.1 Summary of scenarios and overall objectives
The development of the L-mode scenarios in deuterium mimics the development in hydrogen(section 2.5.5.11) and aims at the same final target of 15 MA/5.3 T L-mode plasmas with ~ 50 s
flat-top duration. As in the case of the hydrogen plasmas, the L-mode development will start from
7.5 MA/2.65 T plasmas in deuterium and will progress to 15 MA/5.3 T is steps of Ip/Bt along a q95= 3 - 4 path, with fuelling and power scans being performed at each of the steps. This is expected torequire the Plasma Control System (PCS) for density control and heat load control and thedisruption management/mitigation and avoidance schemes to be retuned from their specifications
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for hydrogen plasmas to adjust to the characteristics of deuterium plasmas. In parallel with thisprogram some elements of the disruption management activities described in section 2.6.3.2 will be
implemented. In addition to the program in L-mode within this operational phase, a precise
assessment of the power threshold to access the H-mode will be performed in deuterium plasmas at
2.65 T. This will include the determination of the density dependence and possible q95 dependencesof the H-mode threshold, in particular at q95=3 where some experiments show increased power
threshold at high Bt compared to that at q95 = 2.7 or 3.3 [Hillesheim, 2016].
The precise determination of the power required to access the H-mode in D plasmas over a range ofconditions is planned at this point, in order to perform the final check (in advance of DT operation)that the baseline power level is sufficient to achieve H-mode DT plasmas at 15 MA/5.3T, required
for the Q = 10 goal, and if not to proceed with the necessary upgrades. Due to the time required to
implement the upgrades, if they are started at this stage, this is likely to lead to a delay of the highIp/Q DT campaign until the upgrades are functional, as only low Ip/Q DT plasmas will be possiblein H-mode. The proposed plan assesses this possible need early in D operations to minimize this
delay.
Assuming that significant relevant experience has been gained in the PFPO-2 hydrogen L-mode
campaign reaching 15 MA/5.3T, the corresponding activities in D (plus H-mode access) describedin this section are estimated to require 32 days of experimental time (plus several additional days
for the disruption management program), divided as follows:

Development of deuterium L-mode operation to 15 MA/5.3 T (26 days)

L-H transition studies in preparation of H-mode operation in D plasmas (6 days)

2.6.3.4.2 Requirements
The experimental plan proposed for this phase considers that the following steps have been
demonstrated:Operation in hydrogen L-mode plasmas up to 15 MA/5.3 T with ~ 50 s successfully
demonstrated;
Disruption avoidance and prediction/mitigation functions have been developed satisfactoryand can be applied without major modifications to deuterium plasmas;
Additional heating systems have been commissioned to full power in D plasmas.

2.6.3.4.3 Risks to the development of deuterium L-mode operation to 15 MA/5.3 T and firstassessment of H-mode access at 2.65 T in D plasmas
The main specific risks for this phase are:

For the L-mode development itself, the main risks are those identified for hydrogen plasmas,
which are related to the use of off-axis ECRH heating for breakdown assist and W control in
the intermediate fields between 2.65 and 5.3 T.

For the H-mode access experiments the main risk is that no mitigation scheme is available to
avoid long ELM-free periods and large ELMs after the L-H transition in D plasmas. At Ip= 7.5 MA this could lead to melting of the W divertor monoblock edges and should be
avoided. These schemes will have been developed already in H plasmas but their application
to D plasmas may not be straightforward due to the different pedestal parameters after the
L-H transition for D than for H plasmas. The mitigation strategy for this risk would be to
first assess L-H access at 5 MA/2.65 T and develop the mitigation schemes in these178

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conditions, which avoid W divertor melting, and then increase plasma current. The
drawback is that q95 varies from 4.5 to 3.0 in this approach. A similar strategy at constant q95
would be possible by starting at 5 MA/1.8 T but this requires pure ohmic breakdown(without ECRH) to be effective at 1.8 T, as the gyrotron frequency will be 170 GHz by thisoperational phase. Either mitigation strategy is likely to require additional operational time.
2.6.3.4.4 Detailed experimental plan for the development of deuterium L-mode operation to
15 MA/5.3 T and first assessment of H-mode access at 2.65 T in D plasmas (32 days)
2.6.3.4.4.1 Detailed experimental plan for the development of deuterium L-mode operation to
15 MA/5.3 T (26 days)
This L-mode operational phase follows the same development of the hydrogen plasmas and is
expected to require 26 days. The deuterium plasma development follows the same Ip/Bt steps
already shown in Figure 2.5-10 and includes scenario #5 to reduce the Ip/Bt step. Table 2-12
describes the scenarios to be explored in this phase; compared to hydrogen plasmas there issignificant additional flexibility for heating of deuterium plasmas because the lower density forunrestricted NBI use with tolerable shine-through loads is much smaller (~ 2.51019 m-3 in D versus~ 4.51019 m-3 in H) and because for low fields (3.3 T) a viable scheme for ICRF heating exists (Hminority). This is expected to make less critical the issue of off-axis ECRH heating for W control in
D plasmas, as more heating mix possibilities exist. Otherwise, the experimental plan is similar tothat for H plasmas (section 2.5.5.11) consisting of fuelling scans, power and power mix scans, etc.,
in the flat-top and ramp-up/down with specific experiments to assess core transport andconfinement properties of D L-mode plasmas. Similarly, there will be experiments to optimize flux
consumption for these scenarios and to control divertor power loads as well as W production and W
accumulation control in the main plasma; this may be already necessary at high additional heatinglevels (see Appendix C).
Table 2-12 – Path to 15 MA/5.3 T (in deuterium)
#

Ip/Bt

2
3

Available heating

7.5 MA/2.65 T
9.5 MA/3.3 T

PECRH (MW)20
20

PICRF (MW)20
20

PHNB (MW)3333

49.5 MA/4.5 T

20

20

33

5

10.5 MA/4.5 T

20

20

33

6

12.5 MA/4.5 T

20

20

33

7

12.5 MA/5.3 T

20

20

33

8

15 MA/5.3 T

20

20

33

2.6.3.4.4.2 L-H transition studies in preparation of H-mode operation in D plasmas (6 days)
To prepare for H-mode operation an assessment of the L-H threshold in D plasmas will be
performed, which is expected to require 6 operational days. This includes the scaling of the H-mode
threshold with density and heating mix at 2.65 T and an assessment of the q95 dependence at q95 ~ 3,
as some experiments show increase H-mode power threshold at this precise value, which is thereference value for Q = 10 operation in ITER. An initial assessment of H-mode access at q95 ~ 5,which is a typical value for steady-state Q = 5 operation in ITER, will also be performed. The
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H-mode access assessment at q95 ~ 3 will be carried out around scenario #2 in Table 2-12 byvarying the plasma current by up to ±10%. The outline of the foreseen experimental conditions
covering the ranges to be explored is summarized in Table 2-13.
As already discussed above the main technical issues of these experiments are:
to achieve the required level of ELM control to avoid long-ELM free periods that can causeedge melting of the divertor W monoblocks and W accumulation in the main plasma. For
this purpose, ELM control by in-vessel coils as well as by the back-up ELM triggering
schemes (pellet triggering and vertical position oscillation) will be applied following theoperational experience in H/He plasmas in PFPO-1 and PFPO-2.
Table 2-13 – L-H transition scenarios (in deuterium)

IP FT (MA)

Bt (T)

Density (fGW %)Heating mix

q95

7.5

2.65

40

NBI+ICRF+ECRH

3

7.5

2.65

80

NBI+ICRF+ECRH

3

7.5

2.65

40

NBI

3

7.5

2.65

40

ECRH+ICRF

3

7.5-10%2.65

40

NBI+ICRF+ECRH

3+10%

7.5+10%2.65

40

NBI+ICRF+ECRH

3-10%

7.5-10%2.65

80

NBI+ICRF+ECRH

3+10%

7.5+10%2.65

80

NBI+ICRF+ECRH

3-10%

4.5

2.65

40

NBI+ICRF+ECRH

5

4.5

2.65

80

NBI+ICRF+ECRH

5

Density scan
Heating mix
scan

q95 scanH-mode accessin q95 of steadystate scenario
density scan

2.6.3.4.5 Diagnostic adequacy for D plasma L-mode scenario studies in FPO
By this stage of the operational program all ITER diagnostics should be available for operation,
including those related to neutron and fast particle measurements. Assuming that no issues appear
related to the functioning of diagnostics until this phase, it is expected that the available diagnostic
set will provide all the measurements required to develop these L-mode scenarios and tocharacterize the plasmas obtained in this phase.
2.6.3.4.6 Deliverables for the development of deuterium L-mode operation to 15 MA/5.3 T andfirst assessment of H-mode access at 2.65 T in D
The main deliverables of this operational phase are:

Demonstration of first L-mode operation in D plasmas at full current and field(15 MA/5.3 T) with ~ 50 s flat-top full integrated plasma scenarios including plasma
position and vertical stability control, power load control, W accumulation control, etc.

Demonstration of the intermediate Ip/Bt steps to be followed to achieve 15 MA/5.3 T Lmode D plasmas along a q95 = 3 - 4 path, which will form the basis of the follow-up
development of the H-mode scenarios in DD and DT;
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Assessment of the effects of gas fuelling and pellet fuelling and additional heating power
level and mix on plasma transport and confinement in ITER L-mode D scenarios up to 15
MA/5.3 T including ramp-up/down phases;

Optimization of flux consumption in D L-mode ramp-up/down phases by using the ITER
systems’ flexibly, in particular, regarding the current ramp rate, and for plasma fuelling andheating in these phases;Assessment of the power required to access the H-mode in 2.65 T D plasmas for a range ofdensities and q95 = 3 – 5, with specific detail on the possible q95 dependence at q95 ~ 3.

2.6.3.5 Development of deuterium H-mode plasmas towards full performance
2.6.3.5.1 Scenarios and overall objectives
The main aim of this first phase of H-mode development in deuterium will be to build a complete
H-mode scenario at half current/half field, which is foreseen to require 40 operational days. This
will have to be reliable, robust to disruptions, MHD stable with controlled ELMs, and will be usedas basis to learn how to use and optimize the additional heating schemes, the tools for ELM control,
NTM stabilization, fuelling and divertor heat load control for H-mode plasmas in ITER. Dependingon the availability of tritium the operational range of H-mode plasma will be expanded towards
higher current and fields in pure deuterium or this expansion will already take place withdominantly deuterium plasmas but with some level of tritium beyond the trace level.
Taking into account the time foreseen to:a) commission the various systems in deuterium plasmas;b) develop the associated L-mode and H-mode access program required before stationary D Hmodes can be studied, and;c) commission the Tritium Plant;
the present plan considers that the expansion of the H-mode operational space in FPO beyond
7.5 MA/2.65 T will be performed in D-dominant plasmas (with T content in the several % level)and not in pure D plasmas in order to optimize operational time, as described in sections 2.6.4.2 and2.6.4.3. If the Tritium Plant were not available, part of this H-mode operational range expansion
would be performed in this phase with pure D plasmas and could require a significant increase of
the operational time allocated depending on the current/ field reached before the Tritium Plant isavailable (up to factors of ~3 - 4).
The main part of the H-mode studies in the first campaign of FPO phase will be carried out in D
plasmas at 7.5 MA/2.65 T, q95 = 3 and will explore the requirements for ELM control with in-vessel
coils building on the findings of PFPO-2. In order to minimize the risk of melting of the divertor
monoblock edges by uncontrolled ELMs in these experiments an option considered is to performthe initial D H-mode characterization and ELM control studies with 5 MA/1.8 T plasmas, where therisk to divertor melting by uncontrolled ELMs is very low (see Appendix C). This option has theadvantage to provide an H-mode development path at q95 = 3 and thus self-similar MHD stability to
7.5 MA/2.65 T plasmas. However, requires the previous demonstration of unassisted breakdown at
1.8 T, as the ECRH gyrotrons will operate only at 170 GHz following the start of FPO. An
alternative approach to reach 7.5 MA/2.65 T H-mode plasmas while avoiding the risks of divertor
melting due to uncontrolled ELMs is to start at 5.0 MA/2.65 T and increase the plasma current
successively; the disadvantages of this approach is that q95 varies from 4.5 to 3 as the current is181

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increased towards 7.5 MA and thus different MHD stability has to be taken into account whendeveloping the scenarios. It should be noted that the restrictions to operation of 5 MA H-modes in
DD in terms of the application of NBI heating are much less serious than in PFPO-2 due to thelower shine-through loads; unrestricted application of NBI only requires plasma densities above
2.51019 m-3, as shown in Figure 2.6-1. This density value corresponds to ~ 0.6nGW for 5 MA
plasmas in ITER, which has been shown to be achievable in L-mode with gas and pellet fuelling[Militello-Asp, 2016]. The lower restrictions regarding the application of NBI, together with the
expected H-mode threshold of 14-19 MW (for 1.8 T and 2.65 T respectively) at the shine-through(minimum) density limit of 2.51019m-3 for D plasmas, is expected to considerably simplify theissues related to the access and sustainment of H-modes at this level of plasma current, whichshould have been extensively studied in PFPO-2.
Similarly, if L-H threshold studies in D plasmas (see section 2.6.3.4) show that small deviations
from q95 = 3 are required to optimize H-mode access, this will affect the exact value of the plasma
current at which H-mode access is performed for the 2.65 T experiments and whether this plasma
current value is kept for the rest of the discharge or later increased to 7.5 MA. Depending on thepossible deviation from q95 = 3 required to optimize H-mode access and sustainment, possible
consequences for achieving ITER project targets at 5.3 T will, then, have to be fully assessed bycomprehensive modelling, benchmarked on the ~7.5 MA/2.65 T performance.

Figure 2.6-1: Normalized NBI shine-through loads (to the value required for unlimited application of
33 MW of NBI power) for 1 MeV D beams in D plasmas versus plasma density.

2.6.3.5.2 Risk to the success of the D H-mode studies in FPO
Most of the risks related to H-mode operation in D plasmas in FPO are linked to the same general
H-mode access/sustainment and ELM control issues previously discussed in PFPO-2 in section
2.5.5.7, namely: H-mode access, sustainment and confinement achievable versus Pinp/PL-H,
conflicting requirements for stationary and ELM power load control, W control and development of182

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ELM control schemes and the avoidance of melting of plasma-facing components. At this point of
the Research Plan it will be known if these risks have materialized or not and whether the
mitigation strategies proposed for PFPO-2 are effective or not. In addition to these risks there are a
few additional ones that are specifically related to deuterium operation:

1.8 T plasma breakdown is not possible without ECRH assist. In this case 5 MA/1.8 T
operation would not be possible and, thus, a q95 = 3 H-mode D plasma with low risk ofdivertor edge monoblock melting would not be available. The mitigation strategy would bebased on two approaches whose suitability depends on the findings during PFPO-2. One
possible approach would be to consider 5 MA/2.65 T plasmas to characterize the H-mode
and then to develop the ELM control strategies with in-vessel coils at q95 = 4.5. Increasing
the plasma current from 5 MA to 7.5 MA would then involve a decrease in q95 towards 3.0
and will require the tuning of the magnetic field perturbation for each step in q 95 to ensurethat ELMs are appropriately controlled to avoid edge monoblock melting, as plasma current
is increased. The other possible approach would be to start the H-mode experiments alreadyat 7.5 MA/2.65 T and make use of the back-up ELM triggering schemes (vertical plasma
oscillations and pellet pacing) to provide ELM control. The possible risk of edge monoblock
melting at 7.5 MA/2.65 T will already have been assessed for H-mode plasmas in PFPO-2,
as well as the efficiency of the back-up ELM triggering schemes to provide ELM control.
The operational experience gained in PFPO-2 will thus determine which the lower risk
approach (for divertor melting avoidance) and fastest approach is to achieve 7.5 MA/2.65 T
H-modes in D plasmas in FPO.

Power threshold and/or the Pinp/PL-H to achieve high confinement are higher than expected
for D plasmas. The D H-mode FPO operational plan is based on the assumption that theexperiments in H and He will provide a more precise evaluation of the power requirements
to access and sustain H-mode D plasmas in ITER. The extrapolation from the power
requirements in H and He to D in ITER will be done on the basis of the understanding
developed in experiments in other tokamak experiments but final confirmation will only beavailable at the beginning of FPO. Given the expected H-mode threshold values for 7.5
MA/2.65 T plasmas, which is in the range of 19-35 MW for <ne> = 0.4 - 1.0nGW, and theavailable heating power of 73 MW, only substantial underestimates of the power required toaccess and/or sustain high quality D H-modes on the basis of PFPO-2 results are likely to
put at risk the Research Plan in this phase up to 7.5 MA/2.65 T. If the power required toaccess and/or sustain high quality D H-modes is significantly higher than expected, thiswould imply that 7.5 MA/2.65 T D H-mode plasma operation would necessarily be
performed at high levels of input power. This would require a high level of availability of
the ITER H&CD systems and the early development and retuning of the control schemes forstationary power load control developed in PFPO-2 already from the start of H-mode
operation in FPO. The implications of these issues, however, are more important for later
phases of FPO when higher Ip/Bt will be explored and higher Pinp will be required. If thisrisk would materialize at the beginning of FPO an upgrade of the H&CD capabilities would
be required to proceed with a significant part of the remaining of the FPO program, as thebaseline capability of 73 MW would not be sufficient to achieve the Q = 10 goal in ITER.
The results of H-mode operation in D plasmas at 7.5 MA/2.65 T would thus provide
guidance regarding the most appropriate upgrades of the baseline H&CD systems to be
implemented in order to ensure the achievement of ITER Q = 10 goals. It should be noted
that, as discussed in Appendix G, depending on the upgrade path required this may imply an

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extension of the D and/or the low-Q DT phase, as the Q = 10 operational phase would notstart until these upgrades have been implemented.
Core-edge integration schemes (ELM control, stationary power load control, W control)developed in PFPO-2 are not applicable for D plasmas. The higher pedestal pressure,higher confinement/higher  H-mode plasmas achievable in D are expected to lead tochanges of the effects of fields applied by the ELM control coils on the plasma, as well as
on W production, transport and exhaust due to changes in plasma response, higher
sputtering yields during ELMs, differences in anomalous and neoclassical transport between
H/He and D, etc. The FPO operational plan assumes that the core-edge integration schemes
developed for H/He plasmas in PFPO-2 will require retuning for D plasmas to account forthese changes but not a complete redevelopment. If this proves not to be true, a significant
additional operational time would be required in this phase to redevelop viable core-edge
integration schemes in 7.5 MA/2.65 T D H-mode plasmas before proceeding to higher
plasma current H-modes in D and DT.

2.6.3.5.3 Details of experimental plan for D plasma H-mode scenarios in FPO (40 days)
The development of the initial deuterium H-mode scenario will, in many respects be parallel to thatof the non-active H-mode development discussed in section 2.5.5.7. Due to the differences in therecycling and atomic properties of deuterium and hydrogen/helium and the expected consequences
for the pedestal plasma, ELM dynamics and overall H-mode confinement, the phenomenology of
H-mode plasmas in deuterium is expected to be quantitatively (and, possibly, qualitatively in some
aspects) to that of hydrogen/helium plasmas.
The experiments will start with 5 MA plasmas and then progressively increase current towards 7.5
MA in steps of 0.5-1.0 MA preferably following a ~ constant q95 approach or, if not possible, aconstant Bt = 2.65 T approach, as discussed above. In each step, as far as ELM divertor power loads
allow without edge monoblock melting, uncontrolled Type ELMy H-mode characterization will be
performed similar to that foreseen in PFPO-2 for heating pulses of similar duration (~ 50 s) andthen the ELM control strategies with RMP fields, W impurity exhaust and those for the control ofstationary power loads with impurity seeding developed in PFPO-2 will be retuned to D plasmas.
The experimental time required for this program is 40 days. It builds-up on the experience gained in
PFPO-2 and will include scans in power levels above the L-H transition, variations in power mixes,
gas versus pellet fuelling comparisons, to determine effects on H-mode confinement, pedestal and
SOL plasma characteristics, etc., as detailed in PFPO-2. This experimental period will conclude
with an integrated demonstration of a 7.5 MA/2.65 T D H-mode plasma scenario which fulfils all
requirements necessary to achieve Q = 10 operation regarding core confinement and core-edge
integration issues such as W impurity exhaust and controlled stationary and ELM transient powerloads to PFCs.
Operation in D H-mode plasmas raises issues that require specific experimental focus ordifferent/additional experiments beyond those detailed for the corresponding phase of PFPO-2.
These include:

Development of optimized additional heating strategies to maximize plasma confinement
while maintaining global MHD plasma stability. The different behaviour of D plasmas
versus H/He plasmas is expected to lead to higher plasma energy/β and thus to different
plasma response to error fields. It is, therefore, foreseen that dedicated experiments will be
required to retune the correction of error fields with the external error field correction coils
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and to optimize the power mix as plasma β increases. At this point it will also be exploredwhether dynamic error field correction by the in-vessel coils is required to optimize H-mode
performance in ITER, given the fact that at these current levels there should be margin in the
ELM control coil system to provide both ELM control and error field correction
functionality (i.e. the current level expected for ELM control is ≤ 45 kAt for a maximum
coil current capability of 90 kAt). Similarly specific tests and optimization of verticalstability control in the presence of ELMs, with abrupt li and p changes, will be included inthis phase as well of the control of NTMs by ECRH/ECCD. In parallel with thesedevelopments the DMS will be retuned to provide the required disruption mitigation as theplasma energy increases.
The additional heating optimization will include assessment of specific issues related to theuse heating systems in D plasmas such as:a) control of the H minority concentration to provide reliable ICRF heating at 2.65 T, which
will have been demonstrated in He plasmas in PFPO-2 and thus with a very different
recycling and pumping behaviour;b) demonstration and optimization of ICRF coupling in the varying edge plasma conditions
expected in D H-mode plasma scenarios;c) re-assessment of the NBI shine-through loads with deuterium NBI in deuterium H-mode
plasmas;d) optimization of the sharing of ECRH/ECCD power between equatorial launcher andupper launcher to provide both W core transport control (W accumulation avoidance) andNTM control, etc.
In particular, if 5 MA/1.8 T plasmas are chosen as the first step of the D H-mode program,specific studies will be required given the expected difficulties to heat these plasmas withICRF (second harmonic minority hydrogen would be required) and also with ECRH (third
harmonic heating of ohmic plasmas may not be effective). Specific strategies to increase theplasma temperature with NBI alone in L-mode plasmas and to access the
H-mode will have to be employed that could be similar to those already applied in the
PFPO-2 phase (i.e. a reduction of the NBI energy and input power) or more advanced (i.e.
reduction in the time-averaged NBI power by cyclic changes of the injected power). The
former would allow the injection of NBI with tolerable shine-through loads at the levels ofplasma density which are expected to be achievable at this current level in ohmic/low power
L-mode plasmas (1.0 - 1.61019 m-3) [Romanelli, 2015; Militello-Asp, 2016]. However, thecorresponding reduction of the NBI power is significant (a factor of ~ 2) and thus reduces
the NBI power to values close to the expected H-mode power threshold at 5 MA/1.8T(10 MW at <ne> = 1.61019 m-3 = 0.4nGW). In this case, robust H-mode operation will
require ECRH heating together with the NBI, which should be effective with third harmonic
heating in these NBI preheated L-mode/marginal H-mode plasmas. The second NBI powerreduction strategy for H-mode access would allow (through the time variation of the NBI
power) to build up the plasma density in L-mode to the level in which unrestricted NBI
operation is possible [Militello-Asp, 2016], thus enabling studies of NBI-alone and NBI +ECRH heating of 5 MA/1.8 T D H-modes to be performed.
Studies of plasma fuelling in the stationary H-mode phase and of H-mode access and exit
phases with pellets and gas fuelling and of the density evolution during confinement andpellet transients. These studies are key to validating predictions of the fuelling efficiency of185

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gas fuelling and recycled neutrals in ITER D and DT plasmas, which is expected to be very
low, to optimize H-mode access and exit through the control of the plasma density in thesephases as well as to optimize pellet fuelling. Some initial experimental evidence on theseissues is expected from PFPO-2 in hydrogen (gas and pellet fuelling) and helium (gasfuelling alone) H-mode plasmas. However, dedicated experiments in deuterium are essential
to provide a quantitative assessment of the above issues with a view to DT operation, due todifferences in edge and pedestal plasmas that affect edge ionization and transport, as well as
in core transport, between H/He and D plasmas. In addition, the restricted operational spacefor H-mode plasmas in hydrogen is likely to restrict the H-mode experiments in PFPO-2 and
it is possible that no significant experimental assessment of these issues can be made inhydrogen H-mode plasmas during PFPO-2. The experiments will explore the range ofstationary H-mode operational conditions and the degree of control of the plasma density in
the H-mode access and exit phases that can be obtained with gas fuelling, pellet fuelling andthe combination of both for a range of currents 5.0 - 7.5 MA, Pinp/PL-H, pellet specifications(size, frequency and injection velocity) and pumping throughput (controlled through shot-toshot variations of the cryopumps’ opening valves). These experiments will perform a full
exploration of the effects of various fuelling and pumping choices on ITER H-mode plasma
scenarios, thus providing essential data to assess the accuracy of the particle transportphysics models and to develop the best H-mode fuelling strategies for access and exit fromburn required for the development of the DT research program.


Exploration of the different scenario options for entrance and exit to high confinement Hmodes, including H-mode entrance in the ramp-up phase and exit in the ramp-down phase,and phasing of the pellet fuelling ramp-up/down versus the heating ramp-up/down. Theseexperiments will build up on the initial experience in PFPO-2 and will expand these results
to relevant conditions with a view to their application to DT plasmas. The experiments will
be carried out in plasmas with a flat-top current of 7.5MA and their main aim is todetermine:a) the evolution of plasma parameters (including li and core W concentration) in these
H-mode access/exit transient phases, and;b) to which level these can be controlled by the heating and fuelling waveforms.
This is essential to start the development of the higher Ip scenarios in DT where the margin
to the H-mode threshold is lower and optimization of -heating is required to have robustaccess and exit schemes to burning plasma conditions that avoid fast back transitions to Lmode and W accumulation [Loarte 2016-2] because, in the higher Ip scenarios in DT, thefine details of access to and exit from burning H-mode plasmas depend on the evolution ofplasma parameters in the core and pedestal in these transient phases (e.g. pedestal plasma
formation for low edge grad-n conditions expected with gas fuelling). The knowledge of
ELM control during these transient phases acquired during PFPO-2 will be applied andretuned to D plasmas. The final goal of these experiments will be the demonstration of the
optimum scheme to provide ELM control with RMP fields throughout the H-mode accessand exit phases up to stationary conditions, both when these access/exit phases take place at
the flat-top current of 7.5 MA as well as during the current ramp-up/down from 5.0 - 7.5
MA in which q95 varies within 3.0 - 4.5. The ELM triggering back-up schemes (verticalplasma oscillations and pellet triggering) will be applied during the transient phases toensure that ELM control is maintained following the guidance of the PFPO-2 results.

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 Specific experiments to demonstrate full edge-core integrated scenario at 7.5 MA/2.65 Twith the required performance for the achievement of Q = 10 in 15 MA/5.3 T plasmas.
These experiments will put together the knowledge acquired in this phase regarding thecontrol of ELM transient loads with 3-D fields, W impurity exhaust and stationary divertor
exhaust by impurity seeding in stationary H-mode conditions and H-mode access/exit
phases and will integrate them to demonstrate a robust H-mode scenario that meets therequirements for Q = 10 operation scaled to 7.5 MA/2.65 T. While this will have been
preliminarily explored in PFPO-2, it is likely that this exploration will have been done onlyfor He H-mode plasmas (given the low margin for hydrogen H-mode operation) and, even if
in hydrogen, at relatively low . Due to the significant changes expected in the SOL,
pedestal and core plasma when operating in deuterium, and the consequences for edge-core
integration issues (ELM control by 3-D fields depends strongly on plasma response,
pedestal W transport and q are expected to depend on neoclassical-related effects, etc.), it isexpected that significant differences will result in the optimization of the various schemes
and strategies for D compared to those applied in PFPO-2. In particular, operation in D
allows an appropriate assessment of the effect that the margin above the L-H transition and have on the overall optimization of the scenarios and the achievable H-mode confinement
providing core-edge integration in ITER as, for D plasmas in these conditions, the ratio of
the input power to the L-H transition threshold power can be varied over a wide range
Pinp/PLH = 1.0 – 4.0 for <ne> = 0.4 - 1.0nGW.
2.6.3.5.4 Diagnostic adequacy for D plasma H-mode scenario studies in FPO
By this stage of the operational program all ITER diagnostics should be available for operation,
including those related to neutron and fast particle measurements. Assuming that no issues appear
related to the functioning of diagnostics until this phase, it is expected that the available diagnostic
set will provide all the measurements required to develop these H-mode scenarios and tocharacterize the plasmas obtained in this phase.
2.6.3.5.5 Deliverables for D plasma H-mode scenario studies in FPO
The main deliverables for this phase are:

Robust 7.5 MA/2.65 T, H98 = 1 deuterium H-mode scenario with integrated fuelling, NTM
and W accumulation control, error field correction, ELM control with 3-D fields from invessel coils to provide acceptable transient power loads and impurity exhaust, stationary
power load control through radiative divertor and with low disruptivity and efficient DMS
when disruptions occur. This provides the basis for predictive modelling and for theoperational development of the 15 MA/5.3 T DT H-mode scenario to be demonstrated in
FPO.



Evaluation of the power level and power mix required to access and sustain highconfinement H-modes in deuterium plasmas. This will provide a high confidence estimate ofwhat would be required for DT plasmas. If the baseline additional heating level is found to
be insufficient for successful Q = 10 operation at this late stage, upgrades of the ECRH
and/or ICRF systems could be implemented to ensure the achievement of the Q = 10 goal,
although the high-Q DT research would need to be delayed until such upgrades are fully
functional.

First evaluation of key plasma behaviour in deuterium H-modes required for the
optimization of plasma scenarios in D and DT plasmas including the degree of control of187

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pedestal and core plasma density, temperature and li by tuning of the heating and fuelling
schemes and waveforms in the transient H-mode access/exit phases and in the stationary
ELM-controlled H-modes.


Confirmation of the ELM control capability that can be expected from ELM control coils in
relevant deuterium plasmas during transient H-mode access/exit and stationary phases overa current range of 5.0 MA – 7.5 MA allowing a high confidence evaluation of thesecapabilities for later high Ip operation in D and DT. If these are found to be insufficient or
marginal, an upgrade of the back-up ELM triggering scheme based on pellet pacing will be
implemented at this stage (e.g. addition of two pellet injectors specially designed for pellet
pacing).

In parallel with the deuterium H-mode scenario development, hydrogenic retention studies,
diagnostic validation and investigations of intrinsic and extrinsic impurity behaviour, etc., will be
performed.
2.6.3.5.6 Progress towards full machine parameters in D H-modes
The main goal of this phase would be to demonstrate a reliable ELMy H-mode scenario atparameters as close as possible to 15 MA/5.3 T with the required performance extrapolable to DT
operation. Based on the considerations above, it is planned to perform this study with plasmas
containing some level of tritium, as discussed in detail in section 2.6.4.1 and these are not includedin the pure deuterium experimental plan in this section. Here we give a short description of the
proposed Research Plan if delays in the commissioning of the Tritium Plant would lead to thesestudies to be done in pure D plasmas.
From the full integrated demonstration at 7.5 MA/2.65 T, the range of plasma currents and fields
would be expanded towards 15MA/5.3T along an approximately constant q95 ~ 3 line, similar to the
L-mode development plan (Ip = 9.5, 12.5 and 15 MA) in Figure 2.5-10. The precise value of current
and fields to be used in the intermediate steps will be the subject of studies during the development,
as they are expected to be determined by the need of central heating by ECRH and ICRF to avoid
W accumulation and the need of off-axis ECRH/ECCD to control NTMs, etc. Given the expected
H-mode threshold in deuterium plasmas and the available maximum additional heating of 73 MW,
it is likely that the maximum plasma current/field where stationary H-mode operation can besustained in D is ~ 12.5 MA/4.5 T, as shown in Figure 2.6-2 Operation at higher plasma currents
and in marginal H-modes is considered to be risky due to possible difficulties in controlling plasma
behaviour in marginal H-mode conditions and the high risk of high Ip disruptions that this entails.

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Figure 2.6-2: Dependence of the L-to-H power threshold, PL-H, in deuterium plasmas versus current, Ip.
The available heating power with central deposition profiles ( < 0.5) for operation at q95 = 3 and
Bt = 5.3 T is also shown.

For each current level a similar set of experiments to those at 7.5 MA/2.65 T is foreseen. This phasewould conclude with a demonstration of an integrated H-mode scenario at each current level. The
total number of days foreseen for these studies would be 120 days but this is not included in theplans for pure deuterium operation described earlier (the experiments are planned to be done
already containing tritium). This time takes into account the more complex development of
scenarios as the plasma current increases. This is due to the increasing deleterious effects of
disruptions and the need to tune the ECRH and ICRF systems to provide both central W
accumulation control and NTM control as the toroidal field varies with increasing plasma current,
because the deposition profiles for ECRH and ICRF are dependent on this.
2.6.3.6 Edge physics (including heat loads) and PWI studies in D plasmas
2.6.3.6.1 Wall conditioning
During the FPO phase, only D-ICWC is of relevance, either at half or full field with
f/Bt = 7.0 – 10.5 MHz/T, depending on the IC frequency. As in the PFPO phases, the efficiency of
D-ICWC for the removal of medium Z impurities can be further evaluated in FPO whenevercleaning discharges are required for programmatic or operational needs. Dedicated trace-T removal
experiments using isotopic exchange with D-ICWC can be also executed in the D phase of FPO(analogous to possible H, D trials in PFPO-2 – section 2.5.5.12.1). Gas balance will be performed,
as before using torus pressure measurements, RGAs and analysis by gas chromatography of the gas
released from the partial cryopump regeneration (to ~100 K). Once experience gained on theoperational parameters to be used (pressure, power, duty cycle and poloidal fields), D-ICWC can beapplied to effectively control the T-inventory build-up in the vacuum vessel, expected to be due
principally to Be co-deposition in the divertor.
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Extrapolations from current devices, in particular the JET-ILW, show that isotopic exchange in
D-ICWC discharges could remove an amount of T comparable to that of the estimated retention per
nominal DT pulse [Brezinsek, 2013-1; Shimada, 2011] (section 2.2.8). However, this may only holdif these discharges are regularly operated before too thick T-rich co-deposits accumulate. Only inthis case will ICWC be able to contribute to mitigation of the T inventory build-up (the so-called‘good housekeeping approach’ [Counsell, 2006]).
2.6.3.6.2 Fuel retention and material migration
By the time pure deuterium operations begin, fuel retention measurements by gas balance will bewell established and operations with pure D will increase the measurement accuracy, allowing
estimates obtained during PFPO phases to be further refined. In particular, characterization of D
retention during H-mode operations at high power with ELM control and a radiative divertor will
expand the existing database. The higher mass of deuterium will lead to an increase of the Be
sputtering from the first wall both in-between and during ELMs. It is therefore important to monitor
material migration regularly. This can be done parasitically during operations through visible
spectroscopy and regular operations of the tritium and erosion monitors. In addition, regular
execution of the reference pulse (Appendix D) will allow for routine monitoring of impurity
sources.
Most of the fuel retention characterization will not require dedicated machine time and will be
performed in parallel with the main scientific program. It will be beneficial, however, to perform a
dedicated fuel removal experiment before the start of trace-T experiments. This will consist in Dremoval by hydrogen ICWC followed by a full bake of the PFCs. Since co-deposits will haveaccumulated during this phase, benchmarking of the fuel removal simulations can be performed
with thicker deposits, allowing validation of the knowledge gained during the PFPO phases.
A similar strategy will be pursued during trace-T experiments where T removal using isotopic
exchange with D can now be developed and validated. An important additional aim of thiscampaign will be to develop/validate the T accounting strategy which relies on measurements
performed in the Tritium Plant for global accounting (including decay), with gas balancemeasurements to determine in-vessel retention and neutron measurements for determination of theburn-up fraction. Performing another fuel removal exercise before the start of DT operations is also
important to validate earlier measurements. It is particularly important in serving to determine thefrequency at which these techniques will need to be employed during DT operations to ensure thatthe in-vessel T-inventory remains below regulatory limits.
2.6.3.6.3 Heat loads and detachment control
With the exception of lower power operation at very low density/divertor neutral pressure, (likely to
be associated with issues of W control), or improper control of plasma position/shape etc., it is onlyin the FPO phases that stationary first wall and divertor target heat loads can attain engineering
limits. By the time FPO begins, power load monitoring and control should have reached a level ofmaturity allowing the progressive increase of input power and current in H-mode which will bemade possible by the switch to deuterium. Progression beyond H-modes at half current and highinput power will in fact not be possible until feedback schemes and robust divertor heat load control
have been fully optimized and shown to be robust and reliable.
High input power (PSOL ≳ 40 MW) deuterium H-mode operation without extrinsic seeding at
7.5 MA/2.65 T (q95 = 3) is expected to produce power flux densities in the strike point regions close
to steady state power handling limits, and considerably above them at low neutral pressure
190

ITR-18-003(Appendix C). Such operation will thus require active detachment control and can be performed atdensities high enough for meaningful application of extrinsic seeding. It is also the point at which
unmitigated ELM energy densities are expected to produce clear W divertor monoblock toroidal
gap edge melting and potential problems with ELM driven W impurity influxes, requiring that ELM
mitigation or suppression be fully developed.
With the progressive increase of plasma current in H-mode which will be possible in FPO, thedatabase of inter-ELM power q,near can be expanded beyond the lower power H-modes accessible
in the PFPO phases, permitting the evaluation of the extent to which the predicted q,near  1/Ip
scaling is preserved with increasing performance. If the scaling is found to apply on ITER up to thehighest currents, the operational window is likely to constrict and detachment control will become
even more stringent as the plasma current is increased.
Heat load characterization should proceed as in the PFPO phases, using first L-modes in deuterium
to expand the PFPO database and applying the detachment control techniques developed in
PFPO-2. The application of 3-D fields at higher current and thus density in L-mode will allow the
study of the effect of divertor dissipation on the perturbed divertor target heat load distribution.
Even 7.5 MA/2.65 T deuterium H-modes are likely to require integrated scenarios coupling ELM
control with impurity seeding, thus requiring the detailed investigation of divertor heat load patterns
in the presence of dissipation and magnetic perturbations, including possible rotation of theperturbation.
Divertor heat load characterization will thus occur along with the scenario development and may
not require much, if any, dedicated experimental time. Some specific experiments will be necessary
to study main chamber power loads, along the lines performed in PFPO-1 and PFPO-2 (e.g. effectof separatrix separation on steady state loading at the top of chamber), though if performed in Hmode there will again be limits to what is possible imposed by the divertor (e.g. the study ofunmitigated ELM impact on the first wall will not be possible if such ELMs are incompatible withdivertor power loads or W influx).
For the first time heat flux control by impurity seeding in H-mode can be rigorously tested underrelevant conditions, including tests of response to loss of impurity injection, the effect of impurity
injection through the main chamber and divertor gas inlets, the detailed dynamics of the
combination of divertor/main chamber fuelling with impurity seeding. A key aspect to examine willbe the relative merits of N2 versus Ne seeding (see section 2.2.5), which, until the advent of higher
power H-mode operation in deuterium, may be difficult to study with respect to extrapolation tofusion conditions as a result of lower pedestal temperatures and plasma densities in lower powerhydrogen H-modes (He H-modes are not useful for the investigation of extrinsic seeding
dynamics/detachment control in view of extrapolation to D or DT). These early H-mode
experiments in D, before the introduction of T, should allow a much clearer assessment of whether
or not, as indicated by plasma boundary simulation for ITER, Ne will be as efficient as N2 as adivertor radiator (section 2.2.5). The use of nitrogen seeding alone will also be important to further
quantify the rate of ammonia formation (section 2.2.4) in more relevant, higher power H-mode
conditions (higher surface temperatures), for which studies will have begun in PFPO-2 and possibly
in PFPO-1. At this stage, first tests of main chamber radiation (Ar) may also be started in
preparation for later higher power fusion plasmas, along with associated feedback control methods.

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2.6.3.7 Trace Tritium H-mode experiments
2.6.3.7.1 Summary of scenarios and objectives
At this stage of the Research Plan, the tritium plant is expected to be able to provide sufficientquantities of tritium to perform trace tritium H-mode experiments in D plasma, i.e. with a muchsmaller throughput (typically more than 100 times lower) than that required for Q = 10 operation in
300 - 500 s burn pulses. This trace tritium campaign is expected to require 12 days of operation andits main aims are to carry out fuelling and particle transport studies of T into D H-mode plasmas,
perform an initial evaluation of tritium retention in H-mode plasmas and to start commissioning thediagnostics required for the DT experimental phase, especially of the 14 MeV neutron diagnostics,
which are key to perform the investigation of core physics processes involving nuclear fuel (tritiumtransport, DT fuel mixture, etc.) in this phase and later in the full DT phase.
The basic scenario to be used for these experiments is the integrated D H-mode scenario at
7.5 MA/2.65 T q95 = 3 developed in section 2.6.3.5 and the main research to be performed willconcentrate on tritium fuelling, particle transport and tritium retention studies with a total
operational time of 12 days.
2.6.3.7.2 Risk to the success of the trace tritium H-mode experiments in FPO
The main risks to the success of these experiments are linked to issues related to the provision oftritium fuel, to its diagnosis and to the associated modelling. The risks associated with H-mode
operation in D plasmas at 7.5 MA/2.65 T are assumed to have been satisfactorily resolved by thisstage:


Late availability of the Tritium Plant. This is not a scientific/technical risk but a schedulingrisk. If this materializes it can be mitigated by pursuing the H-mode research program from
7.5 MA/2.65 T towards 15 MA/5.3 T in D plasmas rather than in DT. In this case theexperiments described in section 2.6.3.5.6 would start at this stage and the program would
revert to the trace tritium experiments as soon as the required tritium throughput is available.

Key diagnostics to determine the concentration of tritium in the confined plasma (e.g.
neutron spectrometer, neutral particle analyzer), in the recycling fluxes (e.g. H and visible
spectrometers/cameras) or retained in the plasma-facing components (e.g. tritium monitor)
are found not to have the sensitivity required to determine trace amounts of tritium (~ 1%)in 7.5 MA/2.65 T H-mode plasmas after commissioning. Depending on the sensitivity
problems and the diagnostics affected, this could be solved by performing the experiments
at a higher tritium level at this stage (~10% T), if this can be provided by the Tritium Plant
or by delaying the experiments until sufficient tritium throughput is available.

Modelling of tritium transport is not yet mature enough at this stage. Modelling of multispecies particle transport in ITER-like plasmas is expected to be very mature at this stage ofthe experimental program. This allows its validation and fine tuning with the trace tritium
experiments and to apply it to the detailed planning of DT experiments. If this turns out not
to be the case there is no risk for the trace T experiments as such but for the follow-up DT
campaign. The DT campaigns will thus have to be fundamentally based on experimental
findings, which will slow down the progress towards maximum fusion performance until
adequate and reliable numerical predictive tools are available.



Trace tritium experiments reveal that tritium retention in ITER PFCs is larger than expectedor tritium removal is less efficient than foreseen. This again is not likely to put at risk the192ITR-18-003

trace T experiments as such, due to the small amounts of tritium fuelled into the plasma but
is a risk for the follow-up DT campaign. Identifying if this risk materializes as an issue as
early as possible in FPO is essential to develop improvements to the foreseen tritium
removal techniques in ITER before routine long-pulse DT operation takes place.
It should be noted that delaying the trace tritium experiments is not a desirable option, as the resultsfrom these experiments are important to understand the behaviour of tritium in ITER H-mode
plasmas and to validate the models on which the detailed planning of the follow-up DT operation
will be based.
2.6.3.7.3 Details of experimental plan for the trace tritium H-mode experiments in FPO (12days)
As describe above, the basic scenario to be used for these experiments is the integrated D H-mode
scenario at 7.5 MA/2.65 T q95 = 3 developed in section 2.6.3.5. This scenario will be used toperform specific experiments to address tritium fuelling transport and retention with tritium
concentrations at the trace level of ~ few %. The diagnostics for tritium measurements in theconfined plasma are designed to measure T concentrations to the 1% level with a time resolution of
100 ms and 20% accuracy [ITER_D_28B39L, 2017]. However, the final capabilities of thediagnostics is dependent on plasma parameters and it may be the case that, for very low values of
the T concentrations, the achievable time resolution is longer than 100 ms if 20% accuracy isrequired. Optimization of the trace tritium level for accurate diagnostic measurements is, therefore,
included as the first item in the experimental plan, which consists of the following steps:
Optimization of the trace tritium concentration level (by T-doped deuterium pellet fuelling)
into 7.5 MA/2.65 T plasmas in order to ensure the highest quality measurements and toallow the commissioning of other diagnostics that will be required for the DT phase beyond
those required to determine the T concentration (e.g. 14 MeV neutron measurements);

For at least two trace tritium levels, fuelling experiments will be performed by injecting thetrace T by gas puffing and T-doped deuterium pellets. DT mix control experiments will be
performed both for stationary H-mode conditions and during the H-mode access and exit
phases. The main goals of these experiments are:- Validate the predictions for T fuelling efficiency by gas fuelling and pellet fuelling
performed on the basis of experiments in D H-mode stationary plasmas and the
optimization of pellet fuelling for efficient core fuelling;- Validate the predictions for core T transport performed on the basis of experiments on
stationary D H-mode plasmas and identify possible effects associated with the different
mass of T versus D on both anomalous transport and neoclassical transport [Loarte, 20162] and ELM control, and to re-optimize the trace T gas/pellet fuelling schemes
accordingly;- Study T fuelling (most likely by pellets) in the transient H-mode access/exit phases with aview to optimize T throughput and to provide robust H-mode access/exit phases in DT
plasmas. These experiments will specifically address the requirements and effectivenessof tritium fuelling to change the DT mix (at the trace level) in the H-mode access/exit
phases by varying the time to start/stop of tritium injection with respect to the start/end of
the H-mode for a range of D L-mode densities (pre and post H-mode);193

ITR-18-003- In parallel with the above, experiments to diagnose tritium retention by in-situ
diagnostics as well as by gas balance will be performed. This will be accompanied bystudies to demonstrate in-situ removal techniques between shots such as ICWC.
2.6.3.7.4 Diagnostic adequacy for trace tritium H-mode scenario studies in FPO
By this stage of the operational program all ITER diagnostics should be available for operation,
including those related to neutron and fast particle measurements. The main open issue, as
discussed above, is whether the sensitivity/time resolution/accuracy for the trace T experiments will
be sufficient to draw quantitative conclusions and to validate the T transport models required toplan DT operation. This is most likely to affect the experiments that concern transients rather thanthose related to determination of the time-averaged T fuelling efficiency in stationary H-mode
conditions. It is possible that the time resolution of the T concentration may have to be increased
beyond 100 ms to maintain measurement accuracy for trace T levels and, depending on theincrease, this may affect the T concentration measurements during H-mode access/exit phases andduring particle transport experiments targeted to characterize the post-pellet inwards diffusion of T
in the H-mode stationary phases.
2.6.3.7.5 Deliverables for the trace tritium H-mode experiments in FPO
The main deliverables for this phase are:
First assessment of the T fuelling efficiency in ITER integrated H-mode scenarios (at7.5 MA/2.65 T) and of the capabilities for DT mix control by gas and pellet fuelling both forstationary and transient phases;Essential data on T transport for the benchmarking the numerical models to be used to
predict the schemes which will be required to achieve DT mix control for optimum burnaccess/exit and stationary burn optimization;

Initial assessment of T retention in relevant conditions for DT operation to confirm theevaluation performed on the basis of D experiments and of the effectiveness of between-shot
T removal techniques such as ICWC.

2.6.3.8 Initial development of hybrid/ advanced scenarios
2.6.3.8.1 Summary of scenarios and overall objectives
The overarching objective for this campaign is to optimize non-inductive current drive fraction in
high-, low collisionality integrated q95 = 4 (hybrid) and q95 = 5 (steady-state) scenarios. The resultsfrom this campaign will establish the basis for those in the DT campaign, ultimately achieving the1000 to 3000 s highly or fully non-inductive discharges with significant levels of fusion power
production. Plasma scenarios with q95 = 4 and 5 at both 2.65 T (5.6 MA and 4.5 MA respectively)
and 4.5 T (9.5 and 7.5 MA respectively) will be the basis for these studies. Consistent with thestrategy followed for deuterium H-modes in section 2.6.3.5, the experiments performed at 2.65 T
will most likely be performed in D pure plasmas while those at 4.5 T will most likely be performed
with some level of T (a concentration of at least 10%), taking advantage of the availability of the
T-Plant at this stage. Should the available T throughput not be sufficient at this stage, the 4.5 T
experiments will be carried out in D plasmas.
In order to accomplish these goals, it is first of all necessary to have full heating and current drive
systems operation for ~ 50 s, fast ion diagnostics and current profile measurements, real-time  and194

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current/pressure profile control developed. The DMS should be fully available and functional andstationary H-modes with robust ELM, density control and divertor detachment real-time control
developed for 2.65 and 4.5 T plasmas with <ne>/nGW ≥ 0.5. This implies that these experiments canonly be performed after the corresponding H-mode program in section 2.6.3.5 for 2.65 T and thepossible follow up in DT for 4.5 T in section 2.6.4.3 have successfully demonstrated thesescenarios.
The estimated operational time required for achieving the overall objectives of this phase isdescribed below. Obviously, these estimates depend on the development performed and research
results from other areas in the program such as H-mode scenario development (section 2.6.3.5) andthe commissioning of advanced control schemes (section 2.6.3.3):
 Application of /current/pressure profile control algorithms using target current profiles for
q95 = 4 and 5 at 2.65 T in stationary H-mode conditions: 10 days dedicated plus experiencefrom other experimental areas; Optimization of performance and non-inductive current drive in 2.65 T discharges at q95 ~ 4
to 5 (Ip ~ 4.5 to 5.6 MA): 10 days;
 Optimization of performance and non-inductive current drive in 4.5 T discharges at q95 ~ 4
to 5 (Ip ~ 7.5 to 9.5 MA): 0 to 20 days (the amount of time in D plasmas depends onwhether the corresponding H-mode experiments are performed in D or in D + 10% T).
2.6.3.8.2 Risk to the initial development for hybrid/ advanced scenarios
The main risks in this phase are related to the capability of the ITER actuators and control loops toachieve the required plasma control and the lack of ability to extrapolate the results in PFPO-2 to D
plasmas. In addition, the general risks affecting the development of the Q ~5 scenarios in Appendix
J also apply:

Uncontrolled impurity accumulation (from W and seeding impurity) degrading plasma
performance and current drive: This can be more problematic in this phase than in PFPO-2
due to the higher plasma currents and associated power loads and may require a higher levelof central heating than in PFPO-2, which may conflict with other j(r) and MHD control
requirements. It is important to note that this issue may not necessarily be problematic in thecorresponding DT plasmas with significant -heating, which provides a central powersource, and probably not likely to be found at this stage for the 2.65 T scenarios for whichsimilar plasmas will have been extensively studied in the H-mode program (see section
2.6.3.5).

Higher  operation leads to additional MHD-related phenomena affecting the scenarios:
The  of the D plasmas in this phase will be higher than that in the H/He plasmas in
PFPO-2. This may lead to new MHD phenomenology affecting NB current drive and fast
particle losses (e.g. AEs, enhanced losses with 3-D fields for ELM control due to larger
plasma response [Kurki-Suonio, 2017], etc.). If this is the case, experimental time will be
required at this stage to characterize and control these phenomena and to mitigate theireffects on the plasma.

Heating and current drive flexibility not sufficient to develop the scenarios: Due to theincreasing requirements on the H&CD it may be the case that the sharing of missions among
the various H&CD systems cannot be satisfactorily met (e.g. ECRH/ECCD may be requiredfor core W control, AE control and NTM control, etc.). If this is the case, an assessment of195

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whether this will be an issue for DT or not will be made and, if so, the upgrade required forsuccessful Q ~5 long-pulse operation in DT identified.
Experience developed in PFPO-2 not relevant: i.e. D plasmas in these scenarios might besignificantly different from those in H/He plasmas so that the operational recipes developedin H/He do not apply. As an example, the current ramp-up scenarios developed in H/He
L-mode may not be applicable to D plasmas due to the lower H-mode threshold if the
H&CD power level applied in this phase is higher than this level, as was the case for D andDT plasmas at JET [Söldner, 1999]. In this case the scenarios would need to be redeveloped
at this stage by adjusting heating level, current ramp-up rate, density, etc.

2.6.3.8.3 Detailed experimental plan for the initial development of hybrid/ advanced scenarios(20-40 days)
Given the large number of issues that need to be assessed and demonstrated in the Research Plan
before these experiments are performed, as well as the remaining physics uncertainties of theextrapolation of the hybrid and advanced regimes to ITER and of their integration with operational
requirements, it is not possible to give a detailed plan for the precise conditions for the q 95 = 3.5-6.0(with focus in q95 = 4 and 5) plasmas at 2.65 and 4.5 T that will be explored. Despite this, it ispossible to identify the specific experimental objectives that will be addressed in these operationaldays:
 Assessment of the diagnostic and control capabilities in longer-pulse plasmas towards
developing control techniques for accessing current profiles consistent with steady-state
operation in ITER;
 Establishment of a basis for fully non-inductive plasma performance in ITER including
initial tests of closed-loop feedback capabilities;
 Development and sustainment of q95 = 4 hybrid discharges with H98 ≥ 1 integrated with all
ITER operational scenario requirements including edge-core integration (fuelling, powerloads, impurity exhaust, etc.), MHD control, etc., and evaluation of expected fusion
performance in DT;
 Establishment of (~ 50 s) plasmas with optimized non-inductive current drive fraction
integrated with all ITER operational scenario requirements including edge-core integration(fuelling, power loads, impurity exhaust, etc.), MHD control, etc., and evaluation of
expected fusion performance in DT;
 Assessment of MHD stability (NTM, AEs, etc.) and confinement properties of long-pulse
q95 = 4 hybrid and steady-state plasmas to determine the control requirements for stable
operation;
 Development of disruption avoidance and detection schemes for highly non-inductive
plasmas and demonstration of disruption mitigation, which are likely to be deviate fromthose for q95 = 3 - 4 plasmas. This will include the assessment of the need for RWM control,
which will be explored within the installed capabilities of the ELM control coils (90 kAt,
5 Hz rotation) and the current level margin available beyond that required for ELM control.
For the Ip/Bt levels in these plasmas, this spare current capability for RWM demonstration is
expected to be > 45 kAt for 2.65 T and > 15 kAt for 4.5 T plasmas;
 Assessment of the pedestal characteristic of hybrid and steady-state scenario plasmas and of
the coupled core and pedestal optimization at the expected operating points for hybrid and196

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steady-state operation in DT. This is expected to require optimization of the heating power
level and mix, torque input, fast particle content and, possibly, plasma density to determine
the dominant physics process that leads to enhanced plasma confinement above H98 in theseplasma scenarios; Complete the validation of models for ECRH, ICRF and NBI heating and current drive in D
plasmas to enable extrapolation to the corresponding hybrid and steady-state plasmas in DT.
This interaction between simulation and experiment is also key to providing input regarding
the decision to go ahead with an ECRH/ECCD upgrade at this stage, if this is essential for
long-pulse Q ~ 5 DT operation, although this would delay that research until this system isavailable (see details in Appendix G). It is essential that the validation of the individualmodels used for these simulations receive high priority during all phases of the steady-state
scenario development, starting with PFPO-2 and most importantly in this initial stage of
FPO. Success in validating these models will provide a powerful tool for planning
experiments during the later DT campaigns in FPO, and ultimately will provide increased
confidence in the ability to project steady-state scenarios achieved on ITER towards Demo.
In this respect, it should be noted that the 3rd HNB upgrade decision should be taken before
FPO (see in Appendix G). Therefore this capability could be available for this operational
period providing more flexibility than the previous experiments in PFPO-2;
 Determining the experimental behaviour of fast ions and validating the models for energetic
ion behaviour resulting from the fast ion populations arising from all heating systems in D
plasmas will be performed in these experiments. These are different from PFPO-2 because
among other things the HNBs are changed from H to D and the ICRF minority at 4.5 T will
be He3, rather than the H minority used at 2.65 T (see details in section 2.5.3). In particular,
if MHD instabilities are observed, determining the nature of the instabilities observed (e.g.
global TAE modes in the outer region of the plasma) [Pinches, 2015] will be a focus of theexperiments;
 If Alfvén Eigenmodes (AE) become unstable, investigating the effects of the various heating
and current drive systems on these instabilities and of the capabilities of these systems toaffect AE instability by local modification of plasma parameters (since this influences theLandau damping experienced by these modes and changes the spectrum of compressionaland Alfvénic modes) [Lauber, 2015] will be a focus of the experiments. If suitable actuators
are established for AE control then establishment and refinement of AE control techniques
and associated controllers could continue at this stage [Van Zeeland, 2016];
 Determining and validating the effects of ELM control coils on fast ion losses for plasmas
with q-profiles as required for hybrid and steady-state D H-mode plasmas [Garcia-Munoz,
2013] will be performed. This would be complementary to the experiments in FPO for the
inductive scenario in section 2.6.3.5.
2.6.3.8.4 Deliverables for the initial development of hybrid/ advanced scenarios
The main deliverables from this operational phase are:

Assessment of the effect of the different isotopes/species (H/He/D) on hybrid and noninductive scenarios development in H-mode;Establishment of a validated model basis for extrapolation of the D plasma scenarios tolong-pulse and steady-state DT operation;

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First assessment of /j(r)/p(r) control in D H-modes and its consequences for plasma
performance;First assessment of the current drive capabilities of the H&CD systems in D H-mode
plasma at q95 ~ 4 and 5 for two Bt-field values in long-pulse H-mode plasma. This
assessment will lead: to refinement in the application of the available H&CD, to theestablishment of the upgrade requirements for the control system sensors and actuators,
including an ECRH/ECCD upgrade for long-pulse development and/or NTM control andthe need to upgrade power supply for RWM active control;

Development of closed-loop feedback control algorithms for core performance and edgecore integration (fuelling, power loads, impurity exhaust, etc.) control;Demonstration of q95 = 4 hybrid discharges with H98 ≥ 1 integrated with all ITER
operational scenario requirements including edge-core integration (fuelling, power loads,
impurity exhaust, etc.), MHD control, etc.;

Demonstration of long-pulse (~ 50 s) plasmas with optimized non-inductive current drive
fraction integrated with all ITER operational scenario requirements including edge-core
integration (fuelling, power loads, impurity exhaust, etc.), MHD control, etc.

2.6.4

Operations Plan for Deuterium-Tritium (DT) Plasma experiments towards Q = 10

2.6.4.1 Plasma scenarios for the experimental program towards Q = 10 operation
The experiments in DT will start from the 7.5 MA/5.3T D and D+trace-T integrated scenarios
described in sections 2.6.3.5 and 2.6.3.7 and proceed by increasing Ip and Bt in steps along a q95 ~ 3line (with Ip = 9.5, 12.5 and 15 MA plus a possible intermediate step at 13.5 MA at constant Bt)
aiming at a stationary H-mode duration of ~ 50 s. For each of the current levels explored, the Tconcentration will be varied from trace levels to ≥ 60 % to determine the effects on the plasma of
the increasing -heating and the optimum T concentration for maximum fusion power production.
As discussed in section 2.6.3.5, it is expected that stationary H-modes with trace T levels will not be
achievable beyond 12.5 MA/4.5 T due to the baseline heating power installed 73 MW and the D
H-mode threshold level. Above this current level, plasmas with significant T concentrations will be
required to achieve stationary H-modes both because the presence of T decreases the H-mode
threshold and because of the increased heating power provided by the additional -heating. In thisrespect, performing this development with significant fractions of T instead in pure D plasmas canmitigate the risks related to changes in the ECRH and ICRF power deposition profiles, which canbecome less central for some values of the intermediate fields between 2.65 T and 5.3 T. In thesecases, -heating will remain centrally peaked and even for the low Q values (~ 1-2) expected at low
Ip levels, the resulting -heating power will be in the range of 10’s MW and thus sufficient to havean important effect on the W concentration according to ITER modelling in [Loarte, 2016-2],
shown in Figure 2.6-3.
The research program is divided into three main phases:The first main phase concentrates on the characterization of H-mode access in DT and thethorough exploration of DT H-modes and on the solution of operational issues in
7.5 MA/2.65T H-mode plasmas requiring 66 operational days;

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The next phase expands the H-mode operational range in DT plasmas up to 15 MA/5.3 T
concluding with the demonstration of Q = 10 operation for burn lengths of ~ 50 s, which is
expected to require 180 operational days;

The final phase performs the developments to increase the Q = 10 burn length to 300 - 500 s
and thus concludes with the demonstration of the ITER inductive goal; this is expected torequire ~100 additional operational days.

Figure 2.6-3: W density profiles, normalized to the pedestal value, for ITER DT 7.5 MA/2.65 T plasmas
with 33 MW of NBI and 0/20 MW of RF for a range of heating schemes modelled with ASTRA. Core
particle and energy transport assumptions where anomalous transport is negligible are those ofneoclassical transport for DT ions and electrons.

It is important to note that it is not expected to reach the Q = 10 short pulse goal during the first
operational period in FPO of 16 months, which amounts to 415 operational days, due to thesignificant commissioning and operational development activities that will have to be undertakenwith the start of D and DT operations. On the basis of the present Research Plan, it is expected thatby the end of the first operational period in FPO fusion power production at the level of ~ 250 MW
for ~ 50 s and Q ~ 5 will have been demonstrated. The plasma scenario in which this will beachieved depends on the result of the research performed within this initial phase of FPO. It iscurrently expected that this will be either with Ip ~ 12.5 MA and <ne> ~ 1020 m-3 (i.e. <ne> ~ nGW)
or with Ip = 15 MA and <ne> ~ 0.71020 m-3 (i.e. <ne> ~ 0.6nGW) [Polevoi, 2015] for the E
predicted by the ITER-H98(y,2) scaling, as shown in Figure 2.6-4.

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Figure 2.6-4: Fusion gain factor, Q, predicted by different core transport models for 13 MA (left) and
15 MA (right) DT plasmas in ITER plasmas with Pinp = 50 - 53 MW. The predictions with the SBM model(Scaling Based Model) correspond to H98(y,2) = 1.

It should be noted that by this phase of operation, ITER will be equipped with the full contingent of
diagnostics required for DT operation. Therefore, no diagnostic limitation specific issues are
discussed further in the sections below. These could arise if key diagnostics do not provide therequired measurements with the foreseen accuracy or if the measurement requirements prove to beinappropriate in view of the achieved plasma parameters in ITER; i.e. if the achieved plasma
parameters differ substantially from those predicted on the basis of existing knowledge that has
been used for diagnostic design. This will only be known when this stage in the operational plan isreached and at that point appropriate mitigation actions will have to be implemented.
In addition, by this point in the operational campaigns, all the tools needed to characterize fuelretention and material migration will have been validated and a good understanding oferosion/redeposition will have been reached. Measurements would then become part of routine
monitoring to check the power handling capabilities and thermal response of PFCs, ensure that FW
erosion remains within allowable limits and that fuel retention rates are as expected and do notchange with time. One important aspect, as wall fluence and integrated material erosion
accumulates, will be to monitor the occurrence of co-deposit flaking, which is known to occur when
co-deposits become too thick. As mentioned earlier, this would represent a potential source of dust
creation in the machine, and would be the dominant dust production mechanism when the
disruption mitigation/avoidance system has reached its target for mitigation rates.
2.6.4.2 Optimization of DT plasma scenarios at 7.5 MA/2.65 T (66 days)
2.6.4.2.1 Experimental plans for the characterization of H-mode access (6 days)
To prepare for the following DT H-mode operation an assessment of the L-H threshold of DT
plasmas will be performed, which is expected to require 6 operational days. This includes thescaling of the H-mode threshold with T concentration, plasma density and heating mix at 2.65 T at
q95 ~ 3 and an initial study of H-mode access at q95 = 5 in the current ramp-up, will also be
performed.
An assessment of the mass dependence of the H-mode threshold will be completed by exploring Hmode access in DT plasmas up to 50% T concentrations, as soon as the T plant allows, bycomparing with the previous D results. Similar to H/He plasmas, H-mode access in the current ramp200

ITR-18-003

will be explored by accessing the H-mode at ~2/3 of the flat-top level (i.e. 5 MA). These points are
essential for the next development of the H-mode scenarios in DT where the optimization of the DT
mix and the heating mix to achieve a robust H-mode transition is required both for the flat-top andthe current ramp-up phase. The outlines of the foreseen experimental conditions covering the ranges
to be explored are summarized in Table 2-14 for the stationary DT plasmas and in Table 2-15 forthe current ramp-up experiments in DT.
As already discussed above, the main technical issues for these experiments are:
to achieve the required level of ELM control to avoid long-ELM free periods that can causeedge melting of the divertor W monoblocks and W accumulation in the main plasma. For
this purpose, ELM control by in-vessel coils as well as by the back-up ELM triggering
schemes (pellet triggering and vertical position oscillation) will be applied following theprevious operational experience in H/He plasmas in PFPO-1 and PFPO-2 and in D in FPO;

to ensure robust DT mix control for the experiments where the T concentration is scannedby gas fuelling and/or pellets.
Table 2-14 – L-H transition scenarios in DT

IP FT (MA)

Bt (T)

Density (fGW %)Heating mix

q95

7.5

2.65

40

NBI+ICRF+ECRH

3

Density scan

7.5

2.65

80

NBI+ICRF+ECRH

3with 10%T

7.5

2.65

40

NBI+ICRF+ECRH

3

Density scan

7.5

2.65

80

NBI+ICRF+ECRH

3with 50%T

7.5

2.65

40

NBI

3

7.5

2.65

40

ECRH+ICRF

3

Heating mix scanwith 50% T

Table 2-15 – L-H transition scenarios in DT during Ip ramp-up
IP ramp (MA)

Bt (T)

Density (fGW %)Heating mix

q95

5

2.65

40

NBI+ICRF+ECRH

4.5

Density scan

5

2.65

80

NBI+ICRF+ECRH

4.5

with 50%T

2.6.4.2.2 Experimental plans for the exploration of DT H-mode at 7.5 MA/2.65 T (60 days)
Beginning with the 7.5 MA/2.65 T D and D+trace-T plasmas, which will have been extensively
explored and understood by this stage, the tritium content of the H-mode phase of the discharge will
be gradually increased, producing higher and higher levels of heating until the optimum T
concentration is exceeded and  heating decreases. In principle, the modification of plasma
characteristics as the T concentration increases in DT 7.5MA/2.65T plasmas is expected to be
gradual, as the maximum level of -heating will be moderate compared to the external heating
sources (< 20%) and the reduction in the H-mode power threshold from pure D plasmas to ~50/50
DT plasmas is expected to be 25%. On the other hand, new physics phenomena may appear with
the introduction of two isotopes that change the behaviour of the plasma qualitatively. For example,
neoclassical DT transport effects due to the different masses of D and T are expected to cause

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density gradients in the central plasma region to appear, which are not existent in D plasmas, that
can affect W transport [Loarte, 2016-2], as shown in Figure 2.6-5.

15
DT

ne

nw

DD

ne

nw

Temperature (keV)

2.0n/nped

1.5

1.0

0.5

0.0
0.00

0.25

0.50

0.75

r/a

T DT

Te

T DD

10

5

0
0.00

1.00

Te

0.25

0.50

0.75

1.00

r/a

Figure 2.6-5: Plasma density and W profiles (left) and electron and ion temperatures (right) for two
7.5 MA/2.65 T ITER plasmas with Pinp = 50 - 53 MW, one in D and the other in DT. Due to neoclassicaltransport effects associated with the different masses of D and T the density profiles are more peaked for
DT in the central plasma region than for D plasmas and this leads to an increased W peaking in DT
compared to D plasmas.

The specific plan for this phase relies on a variation in the tritium concentration from levels of~10% (well above the trace-T level) to levels in excess of 60-70% and, thus, well beyond the
optimum concentration for fusion power production. Various levels of additional heating andadditional heating mixes will be explored and the impact of the increasing T concentration
core/edge/boundary plasma characteristics will be characterized. An important ingredient for thischaracterization to be possible is the demonstration of the capabilities to control the DT mix, which
will have been developed on the basis of the D+trace-T experimental results. For the highest fusion
powers some helium production will take place, although probably at a level one order ofmagnitude lower than for Q =10, that allow fusion-produced He exhaust experiments to start. Most
likely this will require decreasing the pumping throughput to ensure that the core He concentration
exceeds the minimum measurable of 1%.
During this phase, moderate length pulses, few tens of seconds, are likely to be required to assesslong time-scale stability of established feedback algorithms, such as those required for the controlof the DT mix. Therefore, it is foreseen that the stationary phase of these plasmas is at least ~ 50 slong (i.e. ~ 25 E). The final part of the experiments will concentrate on the optimization of fusion
performance and its control with fuelling and additional heating schemes to develop strategies thatwill be used in later experiments at higher plasma currents and higher fusion power wheredevelopment of such schemes is more difficult due to the need to control, simultaneously with thedevelopment, the higher power loads to plasma-facing components. This initial DT H-mode phasewith 7.5MA/2.65T plasmas is expected to require ~ 60 days of operation.
The main experimental activities in this phase are:

Scans of the core plasma T concentration, starting ~10% and going beyond optimum DT
mix, most likely by pellet fuelling, for various levels of additional heating power/additional
heating mix and stationary H-mode plasma density to assess the effects of increasing T
concentration on:
202

ITR-18-003– The L-H power threshold, which is expected to decrease with increasing T
concentration;– Edge pedestal plasma characteristics and ELM power loads (uncontrolled if allowed);
the pedestal pressure and the associated uncontrolled ELM energy losses are expected
to increase with increasing T-fraction;– MHD stability of the core plasma including fast particle effects in DT plasmas (from
NBI, ICRF and fusion-produced -particles);– He concentration in the plasma and its exhaust through uncontrolled ELMs (if allowed)
and controlled ELMs and other collateral effects (e.g. ICRF parasitic resonances,divertor PWI effects, etc.).


The assessment and retuning of control schemes which depend on increasing tritium
concentration and plasma energy (energy confinement is expected to follow T
concentration) and specific scenario issues for DT plasma (preliminarily developed for D
and D+trace-T plasmas):– T fuelling optimization for stationary H-modes (pellet, size, injection velocity, etc.) Hmodes and studies of burn control by fuelling;– Development of robust schemes for robust H-mode access/exit phases including the
optimization of the density waveform before and after the H-mode and the build-up andramp-down of the T concentration towards/downwards the target DT mix in stationary
conditions (e.g. reaching the target DT mix in L-mode and maintaining it through the Hmode, increasing the T-fraction once the plasmas is in H-mode, etc.);– Re-optimization/redevelopment of the schemes to obtain stationary power control by
operating in the radiative divertor regime and evaluation of the consequences for coreplasma confinement, as both W impurity production and transport are expected to beaffected by increasing T concentration;– Re-optimization/redevelopment of the 3-D fields applied by the in-vessel coils toprovide ELM control, which is expected to be required given the effects of increasing T
concentration on the pedestal plasma (increasing pedestal pressure and ELM energy
losses). This re-optimization should integrate other scenario requirements such asradiative divertor operation, impurity and helium exhaust and low losses of fast particles
in the applied 3-D fields;– Re-optimization of NTM control schemes, sawtooth control schemes, error fieldcorrection, disruption and runaway mitigation, etc., which may be required due to theincrease in plasma  and changes to the current profile caused by the increased
T-fraction (e.g. lower li driven by a higher pedestal bootstrap, etc.);– Re-optimization of the H&CD schemes to avoid W accumulation with varying D and T
concentrations.

In parallel with these studies, further studies on T retention and removal (e.g. ICWC) will be
performed to quantify T retention and the effectiveness of T removal techniques now in plasma
conditions very similar to those expected in Q = 10 plasmas.
2.6.4.2.3 Risks for exploration of DT H-mode at 7.5 MA/2.65 T
The major risks for the success of this phase are:203

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Predictive and interpretative modelling of fuelling, transport and -particle effects notmature enough for detailed planning of DT operations at this stage. This will require that theexperimental plan progresses on a more empirical basis, which will require moreexperimental time and possibly smaller T concentration steps and wider parameter scans foreach condition.

Sudden non-linear changes of plasma behaviour above a given T concentration. The
experimental plan above is based on the effects of the T concentration being gradual, as
observed in existing experiments. This allows continuous retuning of the application of theactuators and control schemes from one T concentration to the next. If this is not the case,alternative experimental paths would need to be followed. For instance, ELM behaviour
could already be significantly different at ~10 % T concentration compared to trace-T (orfrom one T concentration level to the next). This could render the ELM control strategies
followed in D and D+trace-T plasmas invalid and risk melting the divertor monoblockedges. In this case, the alternative strategy would be to perform the T-fraction explorationsin 5 MA plasmas (at 1.8 T or 2.65 T depending on the D and D+trace-T experience) andthen gradually increase the plasma current up to 7.5 MA. This would avoid the risk ofdivertor melting but would require more experimental time.

General risks to ITER DT operation could materialize already at this stage, even if a priori
not expected, such as large -particle losses caused by fast particle driven modes, high T
retention etc. These are issues that need to be resolved to achieve the Q = 10 goal and if they
appear at this stage, additional experimental time would be required to resolve them before
continuing with DT program at higher plasma currents.

2.6.4.2.4 Deliverables from the exploration of DT H-mode at 7.5 MA/2.65 T
The main deliverables of this experimental phase are:

Determination of the H-mode threshold with varying concentrations of T in DT plasmas at
2.65 T for stationary H-modes with q95 = 3 and during the current ramp-up at q95 = 5;

Robust 7.5 MA/2.65 T H98 = 1 DT H-mode scenario over a range of densities and additional
heating power with integrated fuelling and DT mix control, NTM and W accumulation
control, error field correction, ELM control with 3-D fields from in-vessel coils to provide
acceptable transient power loads and impurity exhaust (including He exhaust), stationary
power load control through radiative divertor and with low disruptivity and efficient DMS,
when disruptions occur. This scenario will be used to benchmark the predictive modelling to
be used to guide the operational development of the rest of the DT program up to
15 MA/5.3 T Q = 10;

Validated and optimized schemes to control the DT mix in H-mode plasmas in ITER
including H-mode access and exit transient phases with minimum T throughput;

Documented experimental behaviour of DT H-mode plasmas including their reaction toheating and fuelling actuators, which is essential to design the burn control strategies andschemes that will be required in the next step of the experimental program;

Operational experience and experimental data for model benchmark of the re-optimizationof the schemes to control ELM transient loads (ELM control coils) and stationary powerloads (fuelling and impurity seeding) with increasing β and Tritium fraction, which will

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allow a faster progress in the next phase of the DT program at higher plasma currents where
both schemes will be routinely required for operation;
First evaluation of the magnitude of T retention and of the efficiency of removal techniques(e.g. ICWC) in H-mode plasma conditions very similar to those expected in Q = 10;

First assessment of nuclear heating of the toroidal field coils in DT H-mode plasmas. This
together with the AC losses during H-mode operation will indicate if the combination ofboth is likely to be problematic for high current/Q operation. If this is the case, theminimization of AC losses will be an important ingredient of the next operational phase.

2.6.4.3 Optimization of fusion power in DT plasma scenarios towards 15 MA/5.3 T Q = 10(~ 50 s) demonstration (180 days)
Starting from the experience gained in the experiments in section 2.6.4.2 with 7.5 MA/2.65 T
H-mode, the operational space of DT H-mode will be expanded towards the final demonstration of
the ITER Q = 10 goal for burn lengths of ~ 50 s. This is expected to require 180 operational days
and will follow a similar route to that in the development of L-mode plasmas towards 15 MA/5.3 T
discussed in section 2.6.3.4. The approach foreseen for the H-mode development follows the sameroute of gradually increasing current and field while keeping q95 ~ 3, namely the steps of
7.5 MA/2.65 T, 9.5 MA/3.3 T, 12.5 MA/4.5 T, 12.5 MA/5.3 T and 15 MA/5.3 T. The values of thestep ladder to be followed in DT H-modes will be chosen on the basis of those finally found to be
optimum for L-mode, which may be different from those above. Operation along this path makes
use of the flexibility of the ITER H&CD systems to maintain power deposition in the central part ofthe plasma r/a < 0.5, which is required both for plasma performance and to prevent W
accumulation, which should have been demonstrated in L-mode by this stage. In this respect, as
already mentioned before, the increasing production of -heating in these experiments is at a levelthat it is predicted to have a significant effect on W transport and may reduce the demands
regarding the central localization of the deposition profiles of ECRH and ICRF for intermediate
fields between 2.65 and 5.3 T. If this is not sufficient and W accumulation would occur for some of
the Ip/Bt steps above, an alternative route for optimum central ECRH/ICRF deposition would be
followed when increasing Ip, e.g. increase of Ip and constant Bt = 5.3 T, for which q95 would vary.
Plasma operation in this phase will make routine use of the strategies and schemes developed in theexperiments in section 2.6.4.2 for DT mix control, divertor power load control, ELM control, etc.,
because at these current levels the uncontrolled stationary and transient loads are already notcompatible with the ITER PFC lifetime or power handling capabilities, as discussed in Appendix C.
The experiments will consist of steps in Ip/Bt, additional heating power and T concentration until
the Q = 10 goal is achieved and 500 MW are produced in a controlled way.
As mentioned in section 2.6.4.2, in this phase a wide range of T concentrations will be explored at
the lower current levels where DT H-mode can be sustained by additional heating alone (no needfor -heating) while for higher plasma currents only high T concentrations will be explored. As
mentioned in the D H-mode (section 2.6.3.5.6), the expected H-mode threshold for 12.5 MA/4.5 T
plasmas at <ne> = 0.85nGW (the same adopted for Q = 10 operation) in pure D plasmas is 67 MW
compared to an additional heating power of 73 MW so that stationary H-mode operation and low
T- fractions will be limited to low densities near the minimum H-mode threshold of <ne> = 0.4nGW
where Pinp/PLH ~ 1.6.
The experimental strategy adopted streamlines the development of high current H-modes towards
DT by merging the development that was separated in D and DT plasmas in previous versions of205

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the ITER Research Plan. It should be noted, however, that this relies on the tritium plant being able
to deliver an increasing T throughput during this phase to match the increasing T needs associatedwith high T-fraction/high Ip operation.
Within this experimental strategy, the experimental results obtained for 12.5 MA/4.5 T plasmas at<ne> ~ 0.5nGW will be crucial to determine the optimum way to proceed towards higher Ip/Bt in thenext step. These plasmas will be gas-fuelling dominated and are, thus, expected to have relatively
low density gradients in the pedestal. Present estimates with integrated modelling, assuming that the
reduction of transport in the edge transport barrier is not affected by density gradients, indicate thatthe fusion performance that can be achieved in these conditions can be significant (Q > 3) [Polevoi,
2015; Romanelli, 2015] as shown in Figure 2.6-6 , which is due to the high pedestal temperaturesthat can be sustained in these conditions within the calculated edge MHD stability limits. If this isconfirmed by experiments, it would open a route to access high-Q operation in ITER of much easier

Figure 2.6-6: JINTRAC modelled gas fuelled DT
plasmas for a range of plasma currents and withPNBI = 33 MW and PECRH = 20 MW with the GLF23
core transport model. Note the low density
gradients in the pedestal and the high pedestal andoverall plasma temperatures and achieved Q for
these DT H-mode plasmas.

development (i.e. with a more gradual increase of -heating with increasing current), by operatingat <ne> ~ 0.5nGW as the current increases, than the conventional one based on pellet fuelling and<ne> ~ 0.85nGW. In this <ne> ~ 0.5nGW approach Padditional-heating/PLH ≥ 1.7 all the way to
15 MA/5.3 T and, thus, the H-mode plasmas are more controllable with the additional heating andmore weakly dependent on the optimization of the -heating to ensure robust H-mode operation.
In this approach, access to <ne> ~ 0.85nGW Q=10 operation at 15 MA would be undertaken
starting from <ne> ~ 0.5nGW, Q ~5, 15 MA plasmas by increasing gradually the density by pellet
fuelling at this stage. Indeed, an initial phase of <ne> ~ 0.5nGW, Q ~5 to build up -heating before
increasing the plasma density to the nominal <ne> ~ 0.85nGW of the Q = 10 scenario at
15 MA/5.3T has been modelled to provide a more robust access and exit phases to/from burning
plasma conditions both in terms of H-mode control as well as for W accumulation avoidance[Köchl, 2014; Motojima, 2015; Köchl, 2016; Militello-Asp, 2016; Loarte, 2016-2], as shown in
Figure 2.6-7(a) and Figure 2.6-7(b). Otherwise, direct access from L-mode to <ne> ~ 0.85nGW,
Q = 10 conditions relies on the use of the full additional heating power of 73 MW in all discharges[Kessel, 2015] because Pmaxadditional-heating/PLH ~ 1.0 for <ne> ~ 0.85nGW at 15 MA/5.3 T.

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a)b)

Figure 2.6-7: (a) Modelled current ramp-up for ITER DT baseline scenario with L-H transition at 15MA.
Slow (red) and fast (blue) pellet density ramp after the transition. From top to bottom: plasma current,
additional heating level, line average density, Q and maximum divertor power flux and temperature. The
level of 10 MWm-2 is not exceeded in the H-mode access phase. (b) Operational space for the achievementof a transition to high-QDT ~ 10, 15 MA/5.3 T H-mode in ITER for PAUX = 53 MW in terms of the durationof the pellet fuelled ramp to the nominal density (green phase in inset) and of the delay of the pellet fuellingwith respect to the start of the high heating power (blue phase in inset).

The experimental plan foresees performing experiments for each level of Ip/Bt and T-fraction at
various levels of plasma density (at least three <ne> ~ 0.5, 0.7, 0.9nGW) and two levels ofadditional heating one to match N ~ 1.8 and the other to match the expected Pinp/PL-H for Q = 10
plasmas. At each of these steps, the schemes previously developed to avoid/control limiting
instabilities such as NTMs, sawteeth, etc., may need additional optimization time. Similarly, it islikely that at each step the schemes to provide control divertor power loads and ELM control will
have to be retuned. This will become increasingly more complex as the plasma current and fusion
power increases, as the schemes to control divertor power loads and ELMs are likely to affectfusion power production (e.g. through the dilution of the plasma by seeded impurities and theeffects of ELM control on plasma energy) requiring the control loops and the use of actuators to
become increasingly more sophisticated.
Depending on the processes that limit edge MHD stability on ITER plasmas (more ballooning or
more peeling) the pedestal pressure is expected to scale as Ppeelingped ~ IpBt or Pballooningped ~ Ip2[Snyder, 2011; Polevoi, 2015] and thus, if ballooning dominated, an intermediate current point
between at 5.3 T between 12.5 MA and 15 MA may be required to avoid large steps in the edge
plasma parameters. These large steps would make very difficult to ensure that ELM control ismaintained from one Ip step to the next one (i.e. both q95 and edge pedestal pressure would change
significantly when going from 12.5 to 15 MA).

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Other specific tests and studies of the effects linked to disruptions and runaway mitigation withincreasing Ip and/or Bt and plasma energy are addressed separately in section 2.6.3.2 and are notpart of the experiments described here.
In view of the above plan, the main experimental activities foreseen in this phase are:

Increasing the plasma current and field following as much as possible the first two Ip/Bt
points in the step ladder for the L-mode scenarios namely 9.5 MA/3.3 T, 12.5 MA/4.5 T. For
these two current levels the same type of experiments foreseen for 7.5 MA/2.65 T DT
H-mode operations in the previous section will be performed for at least two levels ofadditional heating. This will lead to an increase of the -heating in small steps thus allowing
the tuning of the operational schemes to ensure that the plasmas obtained remain within thecompatibility requirements for ITER operation in terms of power loads, NTM and sawtooth
control, etc.; these will become more complex as the Ip and Bt and plasma heating powerincrease. As mentioned in the description of the overall plan in this phase, with increasinglevels of -particle specific issues related to their effect on plasma MHD stability, -particle
losses due to these effects and to the 3-D fields applied for ELM control as well as He
exhaust may arise at this stage. If this occurs, specific experiments will be required at thisstage to eliminate such possible deleterious effects or to mitigate them before proceeding tohigher current levels. In particular the experimental results obtained for 12.5 MA/4.5 T
plasmas at <ne> ~ 0.5nGW will be crucial to determine the optimum way to proceed towards
higher Ip/Bt in the next step as discussed in detail above.



From 12.5 MA/4.5 T plasmas at <ne> ~ 0.5nGW, the range of low T concentrations that will
be explored will be reduced because significant T-fractions (> 10%) may be required to bothhave an effect on the H-mode threshold and to provide -heating for H-mode sustainment
and W accumulation avoidance Within this reduced T concentration range two routes are
considered feasible to proceed towards 15 MA/5.3 T, Q = 10 plasmas: one based onincreasing Ip/Bt with <ne> ~ 0.5nGW and the other with <ne> ~ 0.7 - 0.9nGW that aredescribed below. It is important to note that in both approaches it is required that scenario
control is maintained in the q95 step (3.0 to 3.6) from 12.5 MA/4.5 T to 12.5 MA/5.3 T.
Given that this is a variation of only 0.6 in q95 and the experience accumulated on H-mode
control with varying q95 (i.e. ELM control by RMP fields, NTM control) at this stage, this isforeseen to be a feasible target but it is likely to require the use of the back-up ELM
triggering scheme of pellet pacing to minimize the risks of this development.

For option 1 (<ne> ~ 0.5nGW) better control of the H-mode scenario can be achieved withthe additional heating, thus making the development less dependent on the optimization of
-heating in the next two higher Ip/Bt steps in the ladder (12.5 MA/5.3 T and 15 MA/5.3 T).
This will allow a wider range of T concentrations to be explored and a more gradualincrease of the -heating, which is expected to ease the progress of experiments. Scenario
control issues are expected to become more complex with increasing current and additional
power levels as well as -power production. At this phase (if already not materialized at
12.5 MA/4.5 T) it is expected that burning plasma physics processes will start to be
dominant allowing both to tackle the issues related to scenario control for Q =10 plasmas aswell as the burning plasma physics issues described in section 2.6.6. Similarly many ITER
systems will start to operate near the maximum of their operational range (i.e. ELM control
coils reaching maximum current capability, ECRH power sharing for NTM and W control
maybe routinely required, etc.) and the need to control the DT mix will increase, as therange of T-fractions over which stationary H-modes can be sustained decreases with208

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increasing current. Every step upwards in additional heating power, T concentration andplasma current from 12.5 MA to 15 MA at (<ne> ~ 0.5nGW) will require the developmentof a fully integrated H-mode scenario before proceeding to the next step, which is expected
to require significant experimental time. In principle, this can be optimized by matching, as
far as possible, the edge power flow level of the 12.5 MA plasmas with high T-fraction withthat of the 15 MA plasmas with lower T-fractions. Once the 15 MA at (<ne> ~ 0.5nGW)
Q ~ 5 scenario has been achieved, the plasma density would be raised in small steps at a
T-fraction of ~50% towards the final goal of <ne> ~ 0.85nGW, Q =10. This strategy is notforeseen to entail special difficulties regarding the sustainment of stationary H-mode
conditions because the H-mode threshold in these DT conditions is expected to increase
with <ne>0.72 while the -heating power increases with ~ <ne>1.5 so that Ptot/PLH increases
with <ne>. By choosing the number of steps to be ~ 5 one can ensure that the increase of heating from one plasma condition to the next one will be of not more than 20% (e.g. fusion
power is expected to increase from 50 to 100 MW and thus 10 MW per step). Advancing
from each step to the next one will require the development of a fully integrated H-mode
scenario to account for the increase of edge power flow due to the -heating, higher He
exhaust required, changes to pedestal parameters and edge stability (increasing nped anddecreasing Tped etc.), increasing plasma  and possibly more strict requirements for NTM
control etc. However, by choosing the density steps judiciously this development could be
simplified, as the changes to plasma parameters, -particle properties, etc., can potentiallybe made as gradual as required. Obviously, phenomena which only occur once a given
physics threshold is exceeded (e.g. fast particle driven instabilities) in this process ofincreasing density and doubling -heating from 15 MA, Q = 5 to Q = 10 will require
dedicated fine-scan experiments for its study and avoidance or for the mitigation of its
effects if required.


For option 2 (<ne> ~ 0.7 - 0.9nGW) careful build-up of the -heating is required to ensurethat appropriate margin of the edge power flow is maintained in each step to provide
stationary H-mode operation; in this case the baseline additional heating is not expected tobe sufficient to sustain stationary H-modes without -heating. This implies the need for
additional dedicated experiments to optimize DT mix control compared to option 1. For thistype of scenario, the key issue for the access and exit from burning plasma conditions is thebuild-up/ramp-down of the -heating with respect to that of the plasma density. A too fastdensity increase after the L-H transition can prevent the plasma from reaching stationary
H-mode conditions as shown in Figure 2.6-7(a). Dedicated experimental time would be
required in this option compared to option 1 to investigate and optimize the scenarios in thisrespect, as this option will most likely rely on H-mode access/exit in the current rampup/down. This will include optimization of the fuelling waveform and of the DT mix and thepower waveform both for access/exit to H-mode in stationary conditions as well as during
the current ramp-up/down (typically at Ip = 2/3I p flat-top). Every step upwards from 12.5 MA
to 15 MA in Ip, <ne>, additional heating power, and T concentration (within the narrow
range that will be explored) will require the development of a fully integrated H-mode
scenario. Given the potential large increase of -heating from 12.5 MA to 15 MA in DT
plasmas at these high densities (a factor of ~ 2) it is likely that at least one intermediate
current point at 13.8 MA will be required in this approach. The development of this option
concludes with the achievement of 15 MA, Q = 10 plasma conditions with <ne> ~ 0.85nGW
for ~50 s, as in the case for option 2.

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Both options have advantages and disadvantages and which one will be chosen will depend both on
the results of physics processes in ITER (pedestal characteristics and achievable H-mode
performance in gas-fuelling dominated plasmas, robustness of stationary H-mode operation with
ELM control for Ptot ~PLH, etc.), as well as on the efficiency of some control schemes versus others(e.g. efficiency and time response of DT mix control, efficiency of ELM control in varying plasma
conditions, etc.). At present it is not possible to judge which one will be finally implemented.
On the other hand, there are many operational and physics issues which are common to both
options. These have to do with the access to burning plasma conditions and the new physics
research that this will allow, as well as the additional scenario integration issues that will have to beaddressed. In both approaches experiments on burning plasmas will be performed at various levelsof -heating to refine schemes for burn control, transient high-Q phases (through additional heating
step-down experiments) will be explored, if a robust scenario can be developed, to determine thebehaviour of burning plasmas fully dominated by -particles, etc.
Regarding scenario integration issues, it is important to note that by this stage of operation many of
the ITER systems may have to operate near the maximum of their operational capabilities (i.e. ELM
control coils reaching maximum current capability, ECRH power sharing for NTM and W control
maybe routinely required, etc.). This is likely to require additional operational time to ensure thatthe scenario requirements can be matched by the hardware capabilities. Examples of this would be:a) optimization of the rotation profiles by varying the power mix for minimum current
requirements for ELM control, if the 12.5 MA experiments show that this would be
marginal for 15 MA operation;b) optimization of the power mix for W control, if the requirements for off-axis NTM control
would not allow sufficient central ECRH power to be available for this mission;c) re-optimization of the current ramp-up/down rate and the stationary H-mode phases, if it isfound that flux consumption and/or AC losses in the PF coils are excessive, etc.
The precise scenario optimization experiments that will be implemented obviously depend on thephysics results obtained and the limitations in the hardware which will be identified at this stage.
The operational plan includes time to perform this type of experiments in order to ensure that theQ = 10, ~50 s goal is obtained within the allocated operational period of 180 days.
With the increasing -heating production and plasma energy, the verification of key systems for
Q = 10 operation will have to be carried out, including: the disruption mitigation system, the effects
of neutrons on diagnostics, possible further calibration and cross-check of neutron diagnostics at
higher neutron production rates as well as the measurements of the neutron heating of thesuperconducting coils with ~ 50 s pulses. This together with the determination of the AC losses is a
key step before proceeding to longer high-Q pulses.
As mentioned above, it is likely that not all of the research under this item will be finished within
the first operational campaign in FPO (i.e. the first 18 months of nuclear operation). The actual
target for the first operational campaign is, therefore, the achievement of several hundred (200-300)
megawatts of fusion power for ~ 50 s with Q ~ 5. The plasma scenario in which this will beachieved will depend on which of the two options discussed above is selected in the final
implementation of the Research Plan.

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2.6.4.3.1 Risk to the optimization of fusion power in DT plasma scenarios towards 15 MA/5.3 T
Q = 10 (~ 50 s) demonstration
By this stage of the operational plan many of the general risks to the achievement of the ITER
Q = 10 goal may materialize. These are general to the project goals and are discussed in Appendix
J. Below we discuss the specific risks that can affect the operational plan in this phase:

H-mode confinement insufficient when integrated with ELM control by 3-D fields with invessel coils (plus pellet pacing) to achieve Q = 10 or ELM control capability is insufficientfor 15 MA due to hardware limitations. This would require a reformulation of the research
program above the plasma current level on which either of these two issues is identified.
Possible mitigation strategies would consist of: a) switching to the hybrid/advanced scenario
development at this stage and proceed to its optimization for Q > 5 and/or b) to proceed todevelop other high confinement regimes ELM-less regimes such as QH-mode or I-mode at
this stage. This can make use of the existing systems (e.g. ELM control coils in non-resonantmode to control edge rotation for the QH-mode) but may also require hardware upgrades tobe implemented. In the latter case, this would be a major strategic decision for the project totake at this stage which will need appropriate justification by dedicated experiments to be
performed in this phase.

Predictive and interpretative modelling of particle and thermal transport through the coreto the edge, and of -particle behaviour required to develop the scenarios and burn control
within them are not mature enough. This will require the progress of the experimental
program to be based mostly on an empirical basis and, thus, slower progress of the researchas the steps in Ip/Bt, additional heating level, T-fraction, etc., may have to be smaller thanstrictly required in order to ensure that the plasma in the scenarios remain integrated with theoperational/hardware constraints and MHD stable to ensure low disruptivity.



Tritium throughput provided by the T-Plant not sufficient at the beginning of theexperimental phase. Although it is expected that the T plant will be able to provide therequired throughput to sustain Q = 10, ~50 s operation by the end of this operational phase,
it may not be able to provide the required throughput at the beginning when 9.5 MA and12.5 MA DT plasmas with high T-fraction are foreseen to be explored. Depending on theactual limitations, various mitigation strategies could be used. One is to develop theseplasma scenarios only at low T-fractions first and then at higher T-fractions (the present planis sequential in Ip and for each Ip step the T-fraction is scanned). The other would be toproceed with the present plan as it is but shorten the burn length by a factor of ~ 2 (25 swould still provide pulses ~ 10E long). The drawback of the latter option is that particle
transport timescales maybe longer than the duration of the stationary H-mode phase andsome experiments would need to be repeated later with longer burn lengths to ensure that thefindings reached with shorter burn lengths remain valid.



Nuclear heating and/or AC losses of the superconducting coils are found to be much larger
than expectations. It is only during this phase that the final levels of several hundreds ofmegawatts of nuclear heating expected in high-Q experiments will be reached, together withhigh current operation in the relevant conditions for high-Q long-pulse scenarios, to allow
these issues to be quantitatively evaluated. It is not expected that such findings would limit
operation at this phase due to the short burn length of ~50 s. However, as already mentioned
above, such findings would require additional scenario development before attempting longpulse Q ~ 5 plasmas. Given that nuclear heating of the superconducting coils can only be
reduced by decreasing neutron production, the additional development of the scenarios
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should focus on reducing AC losses either by optimization of the scenario itself or of thecontrol schemes.
Flux swing provided by the CS/PF system marginal for H-mode operation at 15 MA.
Although the L-mode operation at 15 MA would have already provided some indications if
flux swing would be an issue, it is only in this phase when a realistic evaluation of the flux
consumption of the complete H-mode scenario, including the burn access and burn exit
phases, will be available. If it would be found out that the flux swing is marginal for 15 MA
Q = 10, 300 – 500 s operation, further optimization studies of the scenario will be requiredat this stage. Part of these studies are already included in the research plan for this phase(e.g. access/exit from H-mode at 2/3Ipflat-top) but they should probably be expanded with thespecific target to further decrease flux consumption in the burn access/exit phases.

2.6.4.3.2 Deliverables for the optimization of fusion power in DT plasma scenarios towards
15 MA/5.3 T, Q = 10 (~ 50 s) demonstration
The main deliverables from this operational phase are:

H-mode plasma scenarios over a range of Ip = 12.5 – 15 MA and q95 = 3.0 – 3.6 delivering
hundreds of megawatts of fusion power integrated with all ITER operational requirements
and with demonstrated burn control for periods of ~ 50 s. These are important not only witha view to develop the next phase of the Q = 10 scenario towards 300 - 500s burn length, butalso for the development of the Q = 5, tburn > 1000 s scenarios;

The Q =10 H-mode plasma scenario at Ip = 15 MA and q95 = 3.0 delivering 500 MW fusion
power integrated with all ITER operational requirements and with demonstrated burncontrol for periods of ~ 50 s, essential for the development of the next phase of the Q = 10
scenario towards 300 – 500 s;First results of plasma behaviour in burning conditions allowing the identification of thephysics processes that control this behaviour and the validation/development of models to be
used in the further development of burning plasmas scenarios in ITER and for theinterpretation of the required diagnostics measurements;First results of the TBMs operating at relevant neutron fluxes for high-Q operation (although
at reduced fluences);First assessment of the nuclear heating of the superconducting coils at the levels of several
hundreds of megawatts and of the AC losses for a range of plasma scenarios expanding acurrent range of Ip = 12.5 – 15 MA, which provide key information for the further
development of the Q = 10 scenario towards 300 – 500 s burn length and of the Q = 5,
tburn > 1000 s scenarios.

2.6.4.4 Extension of the Q = 10 scenario towards long-pulse inductive operation (300 – 500 s)(100 days)
After the capability to achieve Q = 10 burning plasma conditions at 15 MA has been demonstrated
for ~50s, the plasma discharges will be developed towards longer durations of 300 - 500s, which is
expected to require 100 operational days. This experimental campaign will not only include thedevelopment of the Q = 10 scenario to long pulse length, but also assessments of long-pulse
operation at a range of Q-values and specific experiments to achieve Q > 10. Obviously, as thepulses become longer additional technical issues will arise such as those related to handling of212

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exceptions by the control system and the control of the plasma termination will become more
complex, as the central solenoid will be operating near its flux swing limit in this phase. While theQ = 10, tburn ~ 50 s will have provided demonstration of scenario integration with many processes
related to power deposition control, -particle physics, thermal and particle transport, DT mix
control, etc., there are processes in ITER which are expected to develop on longer timescales suchas:


MHD-related processes determined by the shape of the current profile, which could lead tothe triggering of tearing modes, Alfvén Eigenmodes, etc., as the current profile changes
shape during the burn phase. This is expected to take from 100 s to 1000 s to reach its final
shape depending on whether the central part of the plasma or the whole plasma is consideredas shown in Figure 2.6-8;

Anomalous transport processes which are influenced by the evolving current profile shape.
The evolving magnetic shear can lead to the onset or change of intensity of turbulent
transport, as shown in Figure 2.6-9 for ITER integrated plasma simulations [Parail, 2013];

Processes controlled by neoclassical particle transport, particularly in the central plasma
region where anomalous transport is low. The values of the neoclassical transportcoefficients both for D, T as well as for W are very low for 15 MA/5.3 T plasmas, as shownin Figure 2.6-10 [Loarte, 2015; Loarte, 2016]. This implies that the typical time evolution
for the DT and W profiles in the central plasma region is of the order of several tens ofseconds and that they may not have reached the fully relaxed conditions in the ~ 50 s burnforeseen for the previous experimental phase;

Other physics processes whose development time could be of the order of tens of seconds.
This could be, for example, sawteeth which are expected to have repetition periods of ~ 50 s
[Hender, 2007], as shown in Figure 2.6-11 and can be stabilized by fast particles in ITER.

Figure 2.6-8: Current relaxation time, as
characterized by the skin time (𝝉𝒔𝒌𝒊𝒏 =𝒂𝟐

𝝁𝟎 𝝆 ,∥

𝒁𝒆𝒇𝒇 = 𝟏. 𝟐,with

Spitzer

resistivity) versus plasma radius for arange of electron temperatures in ITER.

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Figure 2.6-9: From top to bottom: fusion Q, Pfus, Wth, li(3) and line-averaged s/q ratio for ITER JINTRAC
simulations of the entrance to burn including: a reference case with the L–H transition at 15 MA (solid),
and two simulation cases with early L–H transition at 10MA (dotted) and 7MA (dashed), showing thelong-term evolution of the fusion power over time scales exceeding 50s as the current profile shaperelaxes. In the latter case a Greenwald density fraction of 60% was applied to avoid the NB shine-throughlimit after the transition.

Figure 2.6-10: top: Neoclassical diffusion coefficients and pinch velocities for D and T in calculated by
NCLASS and NEO; bottom: Neoclassical diffusion coefficient and pinch velocity for W calculated by
NCLASS. Both D-T and W coefficients are for 15 MA/5.3 T Q = 10 plasmas.

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Figure 2.6-11: ITER simulations using the GLF23
transport model and the complete reconnection
model for the sawteeth. The top frames show theaxial q value, in the middle frames the α-power isthe solid line and stored energy the broken line, andthe bottom frames show the axial temperature[Hender, 2007].

Processes related to long time scale evolution of plasma-wall interactions seen in present
experiments are unlikely in ITER due to the fact that particle/power fluxes are much larger than inpresent experiments and to the fact that the vessel and all in-vessel components are water cooled;
the typical timescales for in-vessel components to reach thermal equilibrium is few seconds.
However, there could be additional issues such as building up of thick co-deposited beryllium
layers for longer pulses that could lead to their overheating and/or detachment for Q = 10,
tburn = 300 – 500 s pulses, which are less likely to occur for the ~50 s pulses, and may pose specificoperational issues when the burn length is extended.
To ensure that these processes and other operational limitations do not increase the likelihood of
disruptions, the extension of the burn length will be done gradually (e.g. in steps of ~ 50 s) andmost likely starting with a lower value of Q. For instance, Q = 5 could be obtained by operation
with <ne> = 0.5nGW for 15 MA/5.3 T or for higher <ne> if T-fractions higher than 50% are used(lower T-fractions would not allow H-mode operation for 15 MA/5.3 T due to the higher H-mode
threshold). Given the lower Q, plasma behaviour can be more effectively controlled with the
H&CD actuators than for Q = 10. Once the Q = 5 scenario is demonstrated for a given burn length,
the Q = 10 scenario with the same burn length will be developed and the process will be repeateduntil the Q = 10 burn length reaches 300 – 500 s. In parallel with these experiments, assessment of
the burning plasma physics processes described in section 2.6.6 and dedicated experiments tocharacterize them will be performed.
This operational period will conclude with a series of experiments to attempt increasing fusion
performance above Q = 10 during transients and during stationary phases taking advantage of the215ITR-18-003

ITER H&CD flexibility and the favourable interaction of fast particle pressure with edge stability
through the increased Shafranov shift (and possibly through a direct influence on anomaloustransport) [Garcia, 2015; Urano, 2016]. By this time of the operational plan the tritium throughput
from the tritium plan will have reach its peak and tritium retention and removal will have become aroutine part of ITER operation. Therefore, it is foreseen that in parallel with the experimental
program, diagnosis of T retention will be done routinely and that T removal schemes between
discharges, overnight and/or over routine maintenance or operational breaks, and/or specific phasesof the discharge designed to remove the trapped tritium will be routinely implemented as part of the
ITER operation cycle.
Consistent with the above global description the main experimental activities during this phase of
the experimental plan will be:Starting at 15 MA/5.3 T, Q = 5 with ~50 s burn, the burn duration will be increased by ~50100 s and a full integrated scenario demonstrated. This includes control of stationary and
ELM-related power fluxes, pumping, DT mix, burn control, He exhaust, MHD stability, etc.,
over the longer timescales.

Once this is achieved, the same scenario will be repeated for Q = 10 and the necessary
retuning of the actuators will be implemented. This sequence will be repeated until Q = 10
for 300 – 500 s is achieved. With increasing burn length flux consumption will be monitored
to ensure that there is enough current margin in the solenoid to provide a robust(i.e. disruption-free) termination of the burn phase and of the discharge as a whole. This is
likely to require re-optimization of the entry-to and exit-from burn phases of the discharge
and possibly of the L-mode ramp-up/down phases that will have to be performed before thelongest burn Q = 10 pulses are attempted.

In parallel, or within the above research activity, specific experiments will be performed.
These will deal with specific control aspects of burning plasmas over long time scales aswell as with the characterization of burning plasma physics phenomena. If deleterious
physics effects, as those described above, appear when the pulses extend over long timescales, specific experiments will be required to mitigate their effects. These could consist ofexperiments to control the evolution of the current profile by local heating and/or current
drive to avoid MHD instabilities, use of H&CD to affect core particle transport and avoid W
accumulation, use of the H&CD systems to modify the fast particle distributions and/or thecurrent profile shape/local plasma parameters to prevent the growth of Alfvén Eigenmodeinstabilities, implementation of specially designed D phases within the high-Q discharges(or special D discharges between high-Q discharges) to remove the retained T, etc.

This experimental period will conclude with attempts to achieve Q > 10 plasmas at
15 MA/5.3 T. This could be done during transient phases which are expected to last several
E (i.e. tens of seconds at most) by gradually decreasing the additional heating power or byusing the flexibility of the ITER H&CD systems to improve fusion performance instationary conditions. One such experiment would consist of the use of He3 minority for
ICRF heating during the burn (this is presently foreseen only for H-mode access/exit in viewof the high He3 consumption expected in ITER). This is expected to increase the level ofcentral ion heating thus increasing core reactivity and Q (up to Q = 1.5 is expected fromthis effect [Wagner, 2010]). Other experiments are foreseen to focus on the optimization of
the favourable interlink between Shafranov shift and edge stability leading to improved
plasmas confinement, which is seen in present experiments. The flexibility of the ITER
H&CD systems can be applied to maximize the central fast particle pressure (e.g. by216

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injecting the beams as close to the plasma centre as possible) and thus Shafranov shift,which can then lead to improved confinement and Q > 10 plasmas in ITER. To which level
this is possible without triggering fast particle-driven MHD instabilities in ITER will have tobe demonstrated. It is also likely that by this stage of the experimental program other ideas
regarding the optimization of H-mode confinement and fusion performance in ITER will
have emerged and these will be exploited as well, before concluding this experimental
phase.
After this operational phase is completed, the experimental program will more strongly focus on the
Q = 5, tburn > 1000 s ITER objective. In this subsequent operational phases, additional inductive
Q = 10 experiments will certainly be performed to assess specific burning plasma issues and,possibly, to provide high neutron fluence for TBM-dedicated experiments. However, in terms ofscenario development, the emphasis will shift towards ITER’s Q = 5 goals and the need to achieve
H98 > 1 confinement with a significant fraction of driven current, as described in section 2.6.5.
2.6.4.4.1 Risks and deliverable from the experimental phase to extend the Q = 10 scenario
towards long-pulse inductive operation (300 – 500 s)
The risks for this experimental phase have either been described in section 2.6.4.3.1 or form part ofthe overall risks of ITER to achieve its goals, which are described in Appendix J.
The main deliverables for this phase are:

Demonstration of robust and integrated H-mode scenarios to achieve Q = 5 and Q = 10
operation for 300 – 500 s, thus achieving the ITER inductive goal;

Determination and control of physics processes taking place in burning plasmas in timescales of hundreds of seconds. This is important both for the achievement of the Q = 10
inductive goal as well as for the further development of tburn > 1000 s, Q = 5 scenarios;First thorough exploration of burning plasma physics processes in ITER and its control
aspects over long timescales;

Demonstration of Q > 10 operation over short and (if experiments are successful) long timescales;

Demonstration of routine operation with acceptable T retention over long pulses andoperational periods at high T throughput;

Testing of the TBMs at high neutron fluxes and fluences.

2.6.5

Operational plan for hybrid/non-inductive DT plasma experiments

2.6.5.1 Summary of scenarios and overall objectives
The data obtained during the first FPO campaign will allow an initial assessment of the mostfavourable path towards long-pulse and steady-state operation. Based on this assessment, the next
step will be to extend the chosen scenarios towards long-pulse operation (q95 = 4 hybrid) up to
1000 s and steady-state operation (q95 = 5) up to 3000 s. While fusion performance will be aconsideration at this point, the overriding consideration will be given to demonstrating long-pulse
operation. Once long-pulse, high-𝛽 N operation has been demonstrated (target 1000 s, high noninductive fraction fNI ≥ 0.6, Q ~ 5). The research program will focus on demonstrating steady-state
fully non-inductive operation (target 3000 s, fNI = 1.0, optimized Q with Q ≥ 2 target).

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The achievement of the final high-Q long-pulse and steady-state goals will require the optimizationof plasma energy confinement together with driven current (by bootstrap current and externalcurrent drive) while ensuring MHD plasma stability and plasma edge compatibility. The
requirements for high energy confinement are less demanding for long-pulse operation (i.e. with
non-inductive current fraction fNI < 1) at high plasma currents. The long-pulse Q = 5, 1000 s goal is
expected to be achievable (and possibly exceeded in terms of burn length) for H98 ~ 1.0-1.1 in ITER
with Ip in the range 13 to 15 MA (q95 ~ 3.5-3) [Polevoi, 2015] and low average densities(<ne> ~ 0.5nG), as shown in Figure 2.6-12 [Luce, 2014]. Long-pulse operation at lower currents~ 11 MA (q95 ~ 4) requires higher plasma density operation and is more demanding with regards toenergy confinement (H98 ~ 1.3), requiring confinement values which are typical for ‘hybrid
scenarios’ at q95 ~ 4 in present tokamaks [Luce, 2014]. Thus the development path of the long-pulse
Q = 5 scenario will have to explore the trade-off between the level of plasma current and energy
confinement that can be achieved in ITER. This will finally determine whether the ‘hybrid’
approach or the low density ‘conventional’ H-mode approach will be more suitable in achieving
this goal. In this section we consider the development path of the hybrid approach towards the
Q = 5 long-pulse goal as this has more synergies with the development path for the Q = 5 steadystate goal, which requires lower plasma current operation with high energy confinement in order toensure fNI = 1.0.

Figure 2.6-12: ITER operational space
diagram for advanced inductive operation at
the nominal ITER toroidal field of
B = 5.3 T with Paux = 50 MW and burnduration of 3000 s. The black curve is theratio of the density to the Greenwald value,
the green curve is the ratio of the loss
power to the predicted L–H threshold
power, and the red curve is the confinement
quality measured by the H-mode scaling[Luce, 2014].

The development path for the Q = 5 steady-state goal (fNI = 1.0) is subject to larger uncertainties
associated with the achievable energy confinement, core density peaking (key to generating corebootstrap current), edge MHD stability (key to providing high energy confinement and pedestal
bootstrap current) and achievable current drive. Figure 2.6-13 shows, as an example, the steady-state
scenarios that can potentially be explored in ITER for 50 MW of additional heating and current
drive power (33 MW NBI plus 17 MW of ECRH) and a core density peaking factor of 1.3,
assuming that the required energy confinement can be achieved. As shown in this figure theachievement of Q = 5 requires H98 ≥ 1.7 and Ip = 8.2-9.1 MA at this power level, while Q = 2.5 canbe achieved for Ip = 9 MA in steady-state for H98 = 1.4 under these conditions. A more detailed
review of the sensitivity of the Q = 5 steady-state scenario to transport, density peaking factor, andheating and current drive power levels and schemes is provided in Appendix E.

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Figure 2.6-13: ITER steady-state operational space with 33 MW of NBI and 17 MW of ECRH and density
peaking of 1.3: a) at prescribed Q=5; b) and c) at prescibed plasma current, Ip =9 MA [Polevoi, 2010].

Therefore, the experimental program in this phase will explore systematic variations of main
plasma parameters (e.g., density, plasma current, applied heating, current profile) and the edge
conditions required for core-edge integration (e.g. divertor radiation for stationary power load
control, ELM control, etc.) will be utilized to optimize performance (core and edge integration)
towards the goal of fully non-inductive operation. During these scans, the control techniquesdeveloped previously in PFPO-2 and earlier in FPO will need to be adjusted in each case to
optimize the control algorithms for the specific plasma conditions. Significant time will be required
during this phase to optimize confinement and real-time MHD stability control in order to sustain
operation near MHD stability boundaries.
At this point in the Research Plan, the main Q = 10 performance objective will have been obtained,
and it is envisioned that an extended operational period of several years will be available for longpulse/ steady-state research. In addition, the required upgrades for the heating and current drive
systems and plasma control systems (e.g. RWM) that have been identified in earlier phases of the
Research Plan will have been implemented and available for use in the development of the
scenarios. The estimated operational time and overall research activities that will be performed inthis period are:
 Application of /current/pressure profile control algorithms using target current profiles instationary DT discharges: 25 days dedicated, synergies with work in other DT
experiments. This research can, in principle, be carried out at any time within the Q = 10
DT development program in FPO, if difficulties in developing that scenario are encountered; Optimization of non-inductive current drive in 2.65 T discharges at q95 ~ 4 (hybrids) andassessment of the influence of isotope (D versus DT): 25 days. This research can, in
principle, be carried out at any time within the Q = 10 DT development program in FPO, if
difficulties in developing that scenario are encountered; Optimization of non-inductive current drive in 5.3 T discharges at q95 ~ 4 (hybrids): 125
days. This research can, in principle, be carried out at any time within the Q = 10 DT
development program in FPO, if difficulties in developing that scenario are encountered; Optimization of performance (with target Q ~ 5) and discharge length (up to 1000 s) in longpulse full-field hybrids (q95 = 4): 100 days;
 Demonstration of 3000 s, q95 ≥ 5 steady-state (fully non-inductive) plasma with Q ~ 2: 100
days;
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 Optimization of 3000 s, q95 ≥ 5 steady-state (fully non-inductive) plasma with target Q ~ 5):
200 days.
The major research focus within this phase will be given to the optimization of the core-pedestal
coupling following on the initial results of FPO on this issue and its integration with the ITER
metallic wall. This will include the assessment of the role of fast particles and plasma rotation on
the virtuous feedback loop between core and pedestal for high- ITER plasmas with a large
fraction of fusion-born fast -particles [Garcia, 2015]. In synchrony with this, significant efforts tounderstand and optimize/control MHD instabilities triggered by fast particles and to reduce fast
particle losses (by MHD instabilities or non-axisymmetric magnetic field effects) are likely to berequired at this stage.
It is important to note that, should the long-pulse, high-𝛽 N operation (target 1000 s, high noninductive fraction, fNI ≥ 0.6, Q ~ 5) development encounter insurmountable difficulties, there
exist alternatives to achieve the 1000 s, Q = 5 goal that do not require a very high inductive current
drive fraction or enhanced confinement (H98 > 1), assuming that the remaining pedestal physics
issues are resolved satisfactorily. These are based on the <ne> = 0.5nGW approach discussed in
section 2.6.4.3 and are a natural extension of the Q = 10 scenario to lower densities and, therefore,
to higher plasma temperatures, implying lower resistive flux consumption. Although this is a lessattractive option from the perspective of fusion energy development towards Demo, it remains a
viable approach to the achievement of the 1000 s, Q = 5 goal in ITER.
2.6.5.2 Risks to operational plan for hybrid/non-inductive DT plasma experiments
These long-pulse and steady-state scenarios require fully integrated capabilities, and the lack of anyof these would affect the accomplishment of this aspect of the project’s mission goals. In addition to
the general risks to the Q = 5 long-pulse and steady-state goals registered in Appendix J, the major
specific risk for this operation phase is that the scenarios developed for D (particularly in the rampup phase) may not be viable for DT due to the different H-mode power threshold, as observed in
JET [Söldner, 1999]. An important issue could arise if the heating and current drive requirements toestablish the required target current profiles at the end of the ramp-up are incompatible with therequired plasma confinement regime in the plasma ramp-up, e.g. if the ramp-up scenarios should
require L-mode confinement to achieve a given j(r), but the required H&CD level during the ramp
were to cause the plasma to enter the H-mode. If this risk materialized, then specific additional
studies, e.g. to prevent access to the H-mode regime to ensure a successful j(r) target at the end of
the current ramp-up, would be required at this stage.
2.6.5.3 Experimental plan to develop DT scenarios for long-pulse hybrid and steady-state
operation with high fusion power production
The ultimate objectives during the DT campaign are: 1) to develop hybrid, q95 = 4 scenarios withpulse lengths of up to 1000 s having high fusion performance (target Q-values of up to 5) and noninductive current drive fractions above 60%; 2) to develop fully non-inductive steady-state
discharges, probably with q95 ≥ 5, having pulse lengths of up to 3000 s (of the order of themagnetic flux diffusion time) with target Q-values of up to 5.
To provide a detailed experimental program for the achievement of these goals to the same level ofthat for Q = 10 (section 2.6.4) is not possible at this stage of the project because such details will
depend critically on the research results of previous operational phases (PFPO-2 and the initial
phase of FPO) and on the upgrades of the H&CD and other control systems (e.g. power supplies for220ITR-18-003

RWM control) that should have been implemented in preparation for addressing these elements of
the Research Plan.
On the basis of the present analysis of long-pulse operation for ITER (see Appendix E for details),
the experimental program needs to develop self-consistent scenarios, in terms of core and edge
plasma confinement (e.g. pedestal MHD stability), and which also satisfy the constraints on coreand global MHD stability, in order to achieve the project’s mission goals. This will need to beachieved by the adjustment and control of the plasma current and pressure profiles using theavailable H&CD actuators and by exploitation of RWM stabilization. These scenarios will have tobe established while satisfying the common requirements for ITER scenarios regarding core-edge
integration (radiative divertor operation, ELM control, impurity exhaust) – these requirements are
expected to give rise to specific issues associated with these novel scenarios due to the high edge
power flows and the lower plasma currents and plasma densities that are expected. Similarly, fast
particle physics and the control of the associated MHD instabilities may require substantial
experimental time, as the proportion of fast particles in these plasma and their contribution to thetotal plasma pressure are expected to be larger than in the intermediate steps towards the Q = 10
goal described in section 2.6.4. This is due to the larger values of the ratio Pheat/<ne> for the longpulse high-Q scenarios compared to conventional H-modes at similar Ip: an increase in Pheat/<ne>should increase N and enhance plasma confinement (by the favourable core/pedestal feedback loop
generating improved confinement) and provide the current drive needed for long-pulse operation.
Although the experimental program to achieve the ITER high-Q long-pulse goals cannot be definedin detail at this stage, it is anticipated that it will consist of the development of a series of DT
plasma scenarios at various levels of plasma current and field towards the final ITER goals. In
many of these plasma scenarios (particularly at low values of plasma current), the value of Q will
be low, so that the influence of -particles on the scenarios themselves is not expected to besignificant. This allows better control of the plasma behaviour by external actuators in thedevelopment of these scenarios but, on the other hand, implies that the effects of significant
-particle heating/fast -particle densities may only be identified rather abruptly in the final stepsof scenario development when the highest Q (~5) will be achieved. This can potentially increase therisk of disruptions at high levels of plasma energy, if core modes are destabilized by the increased
-pressure, and so may require significant scenario redevelopment/retuning in the last steps toachieve the high Q; the outline experimental program discussed below is intended to mitigate such
deleterious effects.
At this stage of the operational program, it is assumed that the target current profiles for thesescenarios, which require tuning of the current ramp-up phase (additional heating power, plasma
density, ramp-up rate, etc.), have already been developed in the previous phases of FPO. If this is
not the case or additional issues appear when the T concentration is increased, additional time will
be required to address these issues; this is not included in the experimental plan below.
The outline experimental plan for the achievement of the ITER long-pulse scenario, Q = 5 goals is
expected to be developed according to the following steps and research objectives:- Development and demonstration of /current/pressure profile control algorithms using target
current profiles in stationary DT discharges (25 days):
The aim of the experiments will be to develop and demonstrate control algorithms with theavailable actuators (H&CD, fuelling, etc.) to control total plasma pressure and current/pressure
profiles in DT plasmas for the range of plasma currents and magnetic fields foreseen in thedevelopment path of the ITER high-Q, long-pulse scenarios. These experiments will be performed
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in plasmas with optimized target current profiles at the end of the ramp-up and will focus on thecontrol aspects themselves, rather than on the optimization of plasma confinement and pulse
duration. This research has synergies with other experiments in the DT program and could, inprinciple, be carried out at any time in the FPO program together with the Q = 10 development.
However, as the actuators for these experiments at this stage of the experimental plan are likely tohave improved capabilities and/or versatility as result of H&CD upgrades, it is considered more
appropriate to carry out these experiments once a substantial research activity towards theachievement of the long-pulse, high-Q ITER goals is launched.
- Optimization of non-inductive current drive in 2.65 T discharges at q95 ~ 4 (hybrids) andassessment of the influence of isotope (D versus DT) in hybrid plasmas (25 days):
The 5.6 MA/2.65 T hybrid plasmas will have been developed during the D phase in FPO, and they
will serve as a basis for assessing the influence of the isotopic mix on discharge development
during the DT phase, which will be very important for the follow-up development at higher toroidal
fields. The experiments will explore a range of D/T ratios (while demonstrating mix control),plasma densities and H&CD mixes, and will aim to optimize current drive and overall plasma
confinement for 2.65 T hybrid plasmas. The low value for the minimum density for acceptable
shine-through loads in DT (~2.51019 m-3) allows unrestricted application of NBI H&CD for<ne>/nGW > 0.6 in these plasmas, facilitating the achievement of very large current drive fractions
even with moderate H-mode like energy confinement, although generating current profiles which
may pose issues for MHD stability (including fast particle effects). This, together with the use of
ECRH for heating and current drive and, possibly, ICRF for plasma heating, will allow systematicstudies of current drive and an initial assessment of the influence of fast particles and plasma
rotation on the enhancement over H-mode confinement that may be achievable in ITER in hybrid
plasmas. Due to the relatively low level of plasma current in these plasmas, it is expected that theedge integration aspects (divertor power loads, W source and transport control, ELM control, etc.)
will pose moderate demands on the development of these scenarios, although especially dedicatedexperiments may be required for the highest levels of additional heating and the lowest densities
explored. Similarly, the effects of -particle heating/fast -particles in these plasmas are expected
to be small due to the low values of Q expected; on the other hand, the influence of fast particles
generated by the heating systems (NBI and ICRF) may be significant.
- Optimization of non-inductive current drive in 5.3 T discharges at q95 ~ 4 (hybrids) (125 days):
These experiments will extend the experience gained in 5.6 MA/2.65 T plasmas to ~ 11.2 MA/5.3 T
plasmas. Although the emphasis in this program will be on optimizing current drive at the levels ofplasma current and toroidal fields required for the high-Q, 1000s pulse goal in ITER, theexperiments in this phase will also perform an initial assessment to determine whether the findings
regarding H-mode confinement optimization for hybrid plasmas at 2.65 T also apply at 5.3 T. It is
important to note that the hybrid experiments for 5.3 T plasmas are expected to be significantly
more complex than those at 2.65 T and thus will require more experimental time due to thefollowing issues:a) Operational experience of hybrid plasmas 11.2 MA/5.3 T in D is unlikely to be available, as
hybrid operation is only foreseen up to 9.5 MA/4.5 T plasmas in D, due to H-mode accessissues. Therefore an intermediate step at 4.5 T with DT plasmas will be required before
exploring these plasma conditions in DT at 5.3 T;b) Due to the higher plasma currents, toroidal fields and additional heating levels, the edge
integration issues (divertor power loads, W source and transport control, ELM control, etc.)222

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will have to be addressed as an integral part of the scenario development; in addition torequiring additional experimental time, this may limit exploration of current drive at thehighest plasma temperatures, emulating conditions expected for Q ~ 5 conditions in this
scenario, which require low density operation;c) The -heating expected in these experiments, although much smaller than the additional
heating, may already be significant (Q ≥ 2) and influence the development of theexperiments themselves (e.g. triggering fast particle MHD instabilities that decrease current
drive efficiency). This may impose restrictions on the range of DT isotopic mixes in whichcurrent drive efficiency can be explored and/or may have implications for the strategy
followed in varying the isotope mix.
- Optimization of performance (with target Q ~ 5) and discharge length (up to 1000 s) in long-pulse
full-field hybrids (q95 ~ 4) (100 days):This experimental phase will concentrate on the achievement of the ITER 1000 s, Q = 5 goal. To alarge degree, the development of scenarios to achieve this goal will be based on the results of theprevious step at 11.2 MA. The experimental results in the 11.2 MA step will indicate the plasma
current level at which the Q = 5 goal is most likely to be achieved; this will depend on the level ofenergy confinement and current drive efficiency achieved, as well as on possible (deleterious)
MHD-related effects. An important aspect of the scenario development and performance
optimization is to avoid large changes in Q from one experimental step to the next to ensure a
gradual tuning of the edge integration and MHD control schemes as plasma fusion performance isincreased. In this respect, it is important to note that the hybrid scenario plasmas can be subject topositive feedback loops between core and pedestal confinement driven by -particle heating/fast-  that can potentially lead to significant increments in Q when the scenario is being
developed towards higher values of Q. Maintaining overall plasma control while sustaining fusion
power production under such conditions will have to be developed and demonstrated. Depending on
the findings of the previous experimental step, two strategies can be considered to reach the Q = 5,1000 s goal:a) To develop a high confinement, MHD-stable (most likely by active control), fully integrated
and controlled scenario with DT plasmas and long burn (>1000 s) at 9.5 MA/4.5 T with thehighest achievable Q, and subsequently to proceed to the development of the equivalentscenario at 11.2 MA/5.3 T to demonstrate the final Q = 5, 1000 s goal; orb) To develop an optimized scenario by initiating the development at 11.2 MA/5.3 T, or evenhigher current (depending on the previous findings), with the necessary adjustment of MHD
control, edge integration schemes, etc., as the burn length and fusion performance increase.
Whether the final experimental path to the Q = 5 goal will follow one or other path, or a
combination of the two, cannot be foreseen at present. As described above, an important operational
issue that may be encountered in such development paths is that of significant changes in Q fromone step to the next due to the occurrence of a positive feedback loop between core confinement
and edge pedestal stability as fusion performance increases. One approach to avoiding thesepotentially large increments in Q is to optimize the scenarios first in T-rich plasmas, taking
advantage of the lower -heating production and lower H-mode threshold in T-rich plasmas than in
50-50 DT plasmas, followed by a progressive change in the DT mix towards 50-50, thus increasing
Q gradually. Note that using D-rich plasmas in such a development path is not a suitable option due
to the higher H-mode power threshold in D plasmas (e.g. for an 11.2 MA/5.3 T D plasma at
ne = 0.8nGW, the H-mode power threshold is ~ 70 MW), which this strongly reduces the flexibility
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in the choice of heating and current drive that can be explored to develop the experimental path tothe desired scenario.
- Demonstration of 3000 s q95 ≥ 5 steady-state (fully non-inductive) plasma with Q ~ 2 (100 days):
The achievement of this fully non-inductive objective requires H98-factors of ~1.2 with Greenwald
factors of ~0.85 (see Appendix E). Achievement of this goal, however, is likely to requiresignificant H&CD upgrades, including the addition of the 3rd HNB as well as an additional 20 MW
of ECRH (case 12 in Appendix E, Table E.2 with q95 = 5.7). The additional NBI injector iscurrently considered to be the most important component of the H&CD upgrade to achieve thisgoal: with the baseline H&CD, the Q value for fully non-inductive operation may fall to 1.3 for H98~ 1.2 (case 2 in Appendix E, Table E.2 with q95 = 7.6).
While the operational experience in FPO and, in particular, the two preceding series of experiments
for hybrid plasmas will have provided important input for this phase, significant issues will remain
open and will have to be addressed as part of this development program. Establishing full noninductive operation under essentially stationary conditions imposes significant requirements on theexternal current drive sources, which also inject substantial heating power and create significant fast
particle densities in the plasma (in particular, NBI). These affect both the current and pressure
profiles, which determine global MHD plasma stability (see Appendix E; internal modes rather than
external ones are more likely in this case given the moderate total plasma ). Fast particle driven
instabilities can also have an important influence on global plasma behaviour in these plasma
scenarios. Significant experimental time will therefore have to be dedicated to the optimization of
the current and pressure profiles via appropriate application of the H&CD and other actuators toachieve and maintain the required plasma conditions over the target duration of 3000 s.
In principle, the level of energy confinement enhancement required for this plasma scenario is inline with expectations for hybrid plasmas in present experiments as well as in ITER, and it istherefore assumed that it will have been demonstrated in the set of experiments that precede thisdevelopment program. The main focus of this phase is to produce an integrated scenario that
demonstrates the steady-state goal at a moderate value of Q (~2). The demonstration of this goal
will require an increase in the fraction of self-driven bootstrap current and, thus, it is expected thatthese experiments will also provide the first empirical insight under ITER conditions into possible
approaches to the optimization of plasma confinement and to the control of MHD stability insteady-state plasmas. Such experience will be essential for the development of the next phase of theexperimental program discussed below. It can be anticipated that the strategy followed for thedevelopment of moderate Q plasmas will be based on the DT hybrid scenario development, and willproceed either by starting from the 9.5 MA/4.5 T scenario, by increasing the field to 5.3 T and
decreasing the current to 9 MA, or from the 11.2 MA/5.3 T scenario, by decreasing the current to 9
MA. If the change in q-profile in following these approaches is found to cause difficulties for thedevelopment of this scenario in DT plasmas (e.g. control difficulties leading to disruptions), an
alternative would be to start the development from the initial scenario at q95 = 5 in D at 4.5 T(7.5 MA/4.5 T) already characterized (see section 2.6.3.8), redeveloping the scenario in DT andthen increasing current and field to 9 MA/5.3 T.
In this phase, edge integration issues are not likely to be significantly more challenging than thosefor the Q = 5 hybrid plasmas that will have been developed previously. These fully-inductive
plasmas will require a high level of additional heating (to increase plasma  and to provide the levelof current drive required to sustain fully non-inductive operation) and will have a lower plasma
density than the hybrid plasmas (due to the lower plasma current), which are in principle more

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challenging for edge integration. This is, however, compensated to a certain degree by the lower
-heating levels targeted in this phase compared to those in the preceding hybrid plasmas.
- Optimization of 3000 s q95 ≥ 5 steady-state (fully non-inductive) plasma with a target of Q ~ 5(200 days)
As discussed in detail in Appendix E, the achievement of this goal requires optimization of the useof the upgraded H&CD systems and other control schemes available in ITER to ensure full current
drive and global MHD stability (including core MHD modes and the control of RWMs) while
maintaining very high energy confinement levels (H98 ≥ 1.6 for q95 = 4.5 - case 9 in Appendix E,
Table E-2 ). Establishing such plasma conditions is expected to require a significant experimental
program, particularly for the achievement of the very high energy confinement levels required. The
details of such a program will depend critically on the experience gained in the development ofhybrid and the fully non-inductive Q ~ 2 scenarios discussed – results from these explorations will
provide essential guidance on potential strategies for combining high-Q (~5) and fully noninductive operation under burning plasma conditions.
It is important to note that if the achievable purely non-inductive and fully integrated plasma
scenario (including all considerations regarding plasma fusion performance, MHD stability andedge integration) cannot finally reach the required fusion performance of Q = 5, other options toachieve the Q = 5 goal during 3000 s can be explored. These correspond to plasmas with a highproportion of current drive (but not 100%) for which the burn duration can extend to 3000 s (verylong hybrid-like plasmas). In these scenarios the plasma current level is larger than for purely noninductive operation but lower than for hybrid plasmas. This reduces the required energy
confinement enhancement (i.e. beyond the reference H-mode level) to achieve the Q = 5 goal; moredetails on such scenarios can be found in Appendix E.
Regardless of whether the 3000 s, Q = 5 plasmas finally developed are purely non-inductive or not,the edge integration issues in these plasmas will have to be addressed simultaneously with thedevelopment of the scenario itself. This will be complex due to the lower plasma densities (lower
edge currents) and high edge power flow (similar to those for the Q = 5 hybrid plasmas) expected inthese scenarios. This integration aspect therefore represents a significant experimental challenge
within this program.
2.6.5.4 Deliverables from the operational plan for hybrid/non-inductive DT plasma
experiments
This operational period, if successful, will demonstrate the achievement of the ITER Q = 5 goals in
long-pulse (~1000 s) and fully non-inductive (‘steady-state’ of ~3000 s) with fully integrated
scenarios. If not fully successful, it should elucidate what the highest achievable values of Q in
ITER are likely to be for both scenarios; this is a very important result for the further developmentof fusion energy. Specific deliverables for this campaign are:

Understanding of the effect of the isotope mix on non-inductive scenario development in
H-modes and on the development of plasmas scenarios incorporating this insight;

Determination of the threshold for, and the consequences of, AE instabilities in longpulse/steady-state plasmas with large -particle populations and the control/mitigation of
their effects as necessary;Assessment of the impact of the -particle population on the scenarios obtained (stability
and confinement);

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Development of fuelling techniques, DT mix control and burn control for long-pulse
operation;

Development of edge-core integration schemes to provide stationary and ELM power load
control compatible with the high level of fusion performance (high confinement) required
for these scenarios;Provision of input to the physics basis for the Demo design and guidance on candidate
operating scenarios.

As described earlier, upgrades of the heating and current drive systems will enable a broader rangeof potential steady-state operation scenarios in ITER, in particular those of relevance to Demodesign/operation. Due to the limited physics basis currently available for these scenarios, it is notstraightforward to anticipate the specific research that will be carried out during this phase.
However, it is expected that the experience gained from present-day devices before ITER operationand from the early phases of ITER operation will provide sufficient information for a detailed
experimental plan, albeit incorporating a range of experimental options, to be developed prior tothis phase. An important output of this program will be an assessment of the viability of, andrequirements for, steady-state operation in Demo; the experiments will be able to address somespecific Demo physics issues such as:Compatibility among high plasma radiative losses (up to the maximum possible radiatedpower fraction), high density (above the Greenwald value) and high fusion performance,
together with integrated stationary power exhaust;

Development of real-time control of integrated high-Q scenarios with a minimum set of
actuators and diagnostics with the power exhaust constraints;



Development of high fusion performance regimes without off-normal transient events,
chiefly edge localized modes and disruptions.

Specific issues that should be addressed within the research activity targeted towards Demo include
the attainable fusion gain in the fully non-inductive steady-state regime, the degree of control
required to maintain steady-state operation, and the associated diagnostic/control requirements. The
additional H&CD upgrades should also allow long-pulse, high gain operation, enabling material
and blanket tests at moderate neutron fluence levels. The latter set of tests would be enhanced
significantly by running a series of highly reproducible steady-state pulses during the latter phase of
ITER operation (see, in particular, section 3.2.2.5.5).
2.6.6

Burning Plasma Physics

2.6.6.1 Opportunities for burning plasma studies in DT plasmas
The range of plasma parameters achievable in ITER provides an opportunity to study the physics oftoroidal burning plasmas in a new, currently inaccessible regime, where a significant population offusion -particles and self-heating are achieved simultaneously. This is particularly true for
operations with Q > 5 when the plasma will be dominantly heated by the population of -particles
created as a result of the DT fusion process. It should be noted that in this regard ITER has
scientific value in providing the validated models that will serve as a basis for Demo design andoperation, even if ITER itself may not provide the explicit experimental demonstration of Demoscenarios.

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The research program in the ITER burning plasma campaigns is likely to evolve as an improved
understanding of key issues is developed within the fusion community during the Construction andearlier Operation Phases of ITER. However, there are certain research issues for which ITER may
be expected to provide the primary contributions to the scientific understanding of the underlying
physics. Some of these issues are shown in Table 2-16 and outlined in sections 2.6.6.1.1– 2.6.6.1.5
below. It should be noted that this list is not meant to be exhaustive, but rather illustrative of theresearch opportunities available on ITER for burning plasma studies in DT plasmas.
Table 2-16 – New phenomena in DT plasmas and expected effects
New phenomena
expected in DT

Expected effects
and potential issues

Significant increaseof plasma heating
power: ~150 MW in
DT vs 73 MW in D

Higher heat loads indivertor, ELM
control, divertor
performance, effecton plasma coreconfinement,
H-mode.
Plasma dilution,
reduction of fusion
power.

He ash accumulation

Self-heating by
fusion -particles
Fast -particle
population in theplasma core

Burn control issues
such as transients infusion power, profile
control, etc.
Additional drive for
Alfvén Eigenmodeinstabilities can cause
loss of -particles,
reduction of heating,
localized heat loads
on the FW; AE
phenomena expected
to be enhanced inlong-pulse plasmas
with higher q; effectof -particles on
MHD phenomena;
additional wall
erosion by ’s.

Experience fromITER H/He and D
phases and present
devices
Power density similar
or higher have beenachieved in present
experiments for shortpulses.

Required additional
dedicated
experimental study
during DT phase
Explore high-Q>10
regimes consistent
with low divertor
load, low tritium
retention.

Externally introduced Investigate HeHe.
profiles in the coreplasma; study of He
exhaust.
Model experiments
Plasma control with
2
with P ~ W possible. minimized auxiliary
power.
AEs driven by NBI
and ICRF minority
fast ions in othermachines as well as
ITER H/He and D
phases.

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Exploration ofstability limits;
investigation ofloss/redistribution vsplasma parameters.

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Isotope effect on
plasma confinement
and edge plasma
performance

Positive effect on
plasma confinementand L-H transition is
expected; scenario
adjustment andoptimization could be
required.

Experience with H,
He, D, DT and T
plasmas in otherdevices and theprevious phases of
ITER.

Extend ρ* scaling
experiments to DT
operation.

The initial phases of DT operation, as detailed in sections 2.6.4.1–2.6.4.3, will begin to addressthese issues associated with -heating and increased power throughput, culminating in Q ≥ 10
operation for a short pulse, of order (10-20) 𝜏𝐸 (several tens of seconds). At this level, many of theessential burning plasma physics issues can be addressed. The subsequent extension of the burn tothe maximum inductive pulse length accessible within the Vs capability of the device in these 15
MA plasma conditions (300 – 500s) will begin to encounter longer time scale phenomena. Broadly
speaking, these include issues of particle transport and exhaust, resistive current diffusion, plasma
material interactions (e.g. tritium retention and impurity generation), and infrequent off-normal
events. While some of the long-pulse development before this phase, may have provided somerelevant experience in most of these areas, the interaction with dominant self-heating and fast (andslowing down) -particles will present new control challenges.
Specific issues to be addressed include:Helium ash build-up and exhaust, and control of the optimum fuel mix;Burn control in the context of time-varying energy, particle (fuel and impurity), andmomentum transport as the q-profile evolves on the resistive timescale;Modification of MHD stability properties due to evolution of the q-profile driven bychanges in the heating profile, including bootstrap current effects, affecting modes whichthen interact with the ’s, including ‘monster’ sawteeth, fishbones, various types of Alfvén
Eigenmodes and NTMs;

Interactions of escaping fast ’s with first wall structures, potentially resulting in sputteringof impurities;

Intermittent or aperiodic relaxation phenomena driven by -heating or instability drive,
potentially leading to off-normal events (e.g. disruptions).

2.6.6.1.1 Fast particle physics and effects of non-axisymmetry
Good confinement of -particles generated in DT reactions is crucial for a fusion power plant.
Single classical orbit -particle confinement is well understood, and -particle loss is predicted tobe relatively low in ITER with corrected toroidal magnetic field ripple. However, a high gradient offast particle pressure in high-Q plasmas can drive collective instabilities that may cause -particle
loss. Collective modes of concern include different types of Alfvén Eigenmodes (AEs), kinetic
ballooning modes, and internal kink modes. In addition to these driven magneto-hydrodynamic
modes, there are also fast particle modes (EPMs) characterized by strong dependence on the fast particle distribution function [Gorelenkov, 2014].
ITER plasma performance also relies on confinement of fast particles generated by auxiliary
heating systems such as NBI and ICRF minority heating. The achievement of H-mode in H/He and228

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D plasmas as well as the transition to high-Q regimes in DT plasmas is critically dependent on theefficiency of heating systems and their ability to deliver full power to the plasmas. Particle energy,pressure, and density of NBI and ICRF generated fast ions are expected to be comparable withthose of -particles in high-Q operation. It is expected that, by the beginning of DT operations,
ITER experiments in the H/He and D phases will already have produced some information on theconfinement of fast particles and other effects of fast particles on the performance of ITER scale
plasmas. High-Q operation in the DT phase will provide a unique opportunity to study fast particle
physics in reactor relevant regimes.
Fast particle effects have been studied in many tokamak experiments using fast ion tails produced
by NBI and ICRF plasma heating, as well as fast -particles in JET [Sharapov, 1999] and TFTR
[Nazikian, 1997] DT experiments. However, direct extrapolation from these experiments to aburning plasma is limited, since the values of characteristic dimensionless parameters and theisotropy of the distribution functions are different. Fast particles in ITER will have much smaller
normalized gyro-radius and have to be confined for a much greater number of bounce periods thanfast particles in present day experiments. Also, the distribution function of fast-particles in a burning plasma will be nearly isotropic, whereas in present experiments, with fast
particle distributions driven by auxiliary heating, distribution functions of fast particles areanisotropic.
Present models predict that the drive due to the -particle pressure, 𝛽𝛼 < 1% in ITER inductive
scenarios, is comparable to that from NBI and leads to marginally unstable toroidal Alfvén
Eigenmodes (TAEs) with n = 15 – 30 in the outer region of the plasma, r/a > 0.4, as shown in
Figure 2.6-14 [Pinches, 2015]. Scenarios with high q95 and, in particular, with reversed magnetic
shear are expected to be more prone to fast particle instabilities. Initial analysis of non-linear
dynamics of EPMs in configurations with reversed shear predicts rapid fast particle redistribution,
albeit with a relatively small global particle loss.

Figure 2.6-144: The sum of -particle and beam drives (red) compared with the primary AE damping
mechanism (ion Landau damping), as a function of normalized radius. The effects of the DT fuel mix (  1  nT nD ) and depletion (   1   nD  nT  / ne ) on the damping are also depicted.

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Fast particles can transiently suppress the internal m/n=1/1 MHD mode resulting in a large
amplitude ‘monster’ sawtooth crash that redistributes fast particles and plasma parameters inside
the affected zone – somewhat larger than the radius of the q(r) = 1 magnetic surface [Campbell,
1988; Chapman, 2007]. The sawtooth crash can induce seed magnetic islands to initiate the growthof neoclassical tearing modes (NTMs). On the other hand, fast particles can destabilize the internal
m/n=1/1 MHD mode through a resonant wave-particle interaction at the magnetic precession
frequency of the trapped fast ions, and result in fishbone oscillations that can also redistribute thefast particles within the central zone.
Given the large number of instabilities and the complexity of the problem, computational codes
predicting -particle driven instabilities in burning plasma are unavoidably approximate andsensitive to the detailed profiles of the scenarios. This makes the development of predictive
capabilities particularly challenging despite ongoing validation exercises against present day
experiments.
Physics research in the area of fast particles in burning plasma will require additional measurements
such as the profile of fusion power (e.g. neutron tomography), fuel composition including D/T and
He/D ratios and impurity concentrations, and MHD fluctuation in the range of Alfvén modes (30300 kHz). If anomalous loss of fast -particles is encountered, then measurements of the energy
spectrum of confined -particles and the first wall flux of escaping ’s will be required in order tounderstand the loss mechanism and to try to eliminate the cause. To avoid localized power loads on
the FW – which is typical for fast particle loss – a monitoring of localized heat loads on the plasmafacing wall will be necessary, especially for the high-Q operation. Whilst visible/IR cameras are
sufficient for machine protection purposes, understanding the physics behind such loss channels
will require dedicated fast ion loss detectors that can resolve more information about the specificdetails of the lost fast ions, e.g. their pitch angle and energy. Such diagnostics are currently underinvestigation for implementation in the baseline, namely the fast ion loss charge-exchange
diagnostic and a fast ion loss detector (see Appendix H).
In addition to non-axisymmetric effects arising from the presence of MHD phenomena, theapplication of 3-D magnetic fields to influence ELM behaviour will also have an impact on fast
particle confinement. Detailed modelling, in which the important effect of the plasma response isretained, has shown that, for the maximum current in the ELM control coils in the baseline scenario(15 MA), an additional fast ion power loading in the divertor region is primarily observed, as
illustrated in Figure 2.6-15 [Akers, 2016]. This should be monitored as the populations of fast ionsare built up during the experimental program and also as various levels of current and relative
phasing are explored in the ELM control coils. Whilst the distribution of fast ion power loading in
the divertor is largely determined by the geometry of the divertor structure, the loads onto the firstwall are determined by the applied spectrum. If these loads reach levels where they are not tolerable
in stationary conditions then it may be necessary to use the flexibility within the ELM control coils
to change the applied field during a discharge in order to move the heat loads around whilst
simultaneously meeting the requirements for ELM control (e.g. by statically rotating the appliedfield or changing the harmonic mix to otherwise modify the heat load footprint and protect the firstwall during the discharge evolution).

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Figure 2.6-15: Fast ion power loading to the ITER under-dome components for 33 MW of 1 MeV
NBI into 15 MA flat-top H-mode with 90 kAt of n = 3 ELM control coil field applied (including then = 6 side-band). The power loads in a) assume that a vacuum approximation holds whilst b) includes
the plasma response [Akers, 2016].

2.6.6.1.2 Self-heating and thermal stability
Significant plasma self-heating by -particles raises the issue of thermal stability of the operating
point for burn. While one can conceive burn control simulation experiments in present devices orthe earlier phases of ITER operation, in which a fraction of heating power is proportional to Wth2 (Pfus in a power plant plasma), thermally stable operating modes for a power plant must be
demonstrated in a plasma heated by fusion reactions. The plasma burn in ITER and Demo are
expected to be globally stable, with operation located near the stable (right) branch of the ignition
curve where the power loss increases faster with temperature than the fusion power [Rebhan, 1997].
An expected increase in fuel dilution by helium ash with rising fusion power is an additional
stabilizing factor. However, the consistency between the plasma pressure profile, the current density
profile (including bootstrap current), and the plasma stability requirements may raise multi-faceted
feedback and profile control issues that could be important, especially for hybrid and steady-state
scenarios in which the power degradation of energy confinement is expected to be weaker than inconventional H-modes [Challis, 2015]. In such a case, a simple global control scheme may not beadequate and more sophisticated control of plasma current and pressure profiles may be required.
Possible actuators for fusion power control in ITER include auxiliary power, fuelling, and impurity
injection. Even at Q ≈ 10, substantial external heating power is still required at the burn point, and acontrol strategy for regulating the fusion power by modulation of Paux seems feasible. Actuators forparticle control include gas puffing and cryogenic pellet injectors. Reduction of the fuel ion density
reduces the reactivity as <nDT>2, but the response to the fuelling source (primarily pellet injection)is on the slower particle confinement time scale. More importantly, reduction in the core density
tends to reduce the radiated power fraction unless additional impurity seeding is provided
simultaneously, thus aggravating the power exhaust challenge to the divertor.

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Conversely, positive excursions in the density to counter a decrease in reactivity are bounded by theproximity to the density limit. The concept based on controlling the fuel mixture (ratio n D/nT) could
be tested on ITER using independently timed pellet injectors for D and T fuelling.
Measurements for the various quantities of interest for control will be available in ITER. The fusion
power (proportional to 𝑃𝛼 ) itself is observable with essentially zero lag by monitoring the flux of
14.1 MeV neutrons. This is an important consideration, since the plasma response to variations inreactivity experiences a lag associated with the –particle slowing down time and the energy
confinement time, typically amounting to a delay of several seconds. Real-time spatial profiles of
the kinetic quantities ne, Te, Ti, and to some extent of radiating impurities, are expected to beavailable on ITER using the foreseen diagnostic set. Real-time measurements of global quantities
such as the plasma current and plasma pressure (𝛽), and of the plasma shape, will be available frommagnetic sensors. Measurements of the fuel mix at the plasma edge, or in the divertor region using
neutral emission (e.g. 𝐷𝛼 /𝑇𝛼 ) seems feasible, as does monitoring of the exhaust gas. Alternatively,
there is the possibility of inferring the relevant core ratios from simultaneous observation of the
2.45 and 14.1 MeV neutrons arising from the D(T,He4)n and D(D,He3)n reactions respectively.
Although ITER’s objective is the achievement of Q ≥ 10, much higher Q values, and even ignitionin ITER, are not precluded within the range of present uncertainties in projections of ITER plasma
parameters. Optimization of operational scenarios with the aim to increase the Q value will be an
important part of ITER research program during the DT active phase. As one can expect, an
increase in Q to very high values will make plasma control more difficult, but it would provide a
unique environment in which to develop plasma control necessary for Demo.
2.6.6.1.3 Macroscopic stability physics and control
Issues of macro-stability research to be addressed during the ITER DT phase operation are centred
on four main topics:MHD activity driven by the -particles, discussed in section 2.6.6.1.1;
Confinement limiting instabilities, exemplified by NTM modes in the core and pedestal
stability and relaxation mechanisms in the edge;
Ultimate pressure (β) limiting instabilities, exemplified by RWM;
Off-normal events, in particular disruptions.

These effects must be understood and dealt with in ITER in order for the facility to achieve its
performance objectives. Moreover, research into these instabilities on ITER is essential in order toprovide data in support of Demo and reactor facilities, either by directly demonstrating relevant
conditions or to validate theoretical extrapolation to parameters not actually achievable in ITER
itself.
DT operation in ITER will provide a unique combination of βα, a/ρα, and vα/vAlfvén for the study of-driven instabilities, especially Alfvén Eigenmodes. While the projected value of β α (<1%) in
ITER is smaller than might be expected in a reactor, for example, the comparable value of the ρα,isotropic source distribution, and self-heated burn dynamics make the ITER data an invaluable
extension to the database in this area by providing a vital testbed for validation of linear and nonlinear theoretical models of these phenomena. The ITER research program in the area of-driven instabilities would therefore incorporate:Validation of linear stability boundaries in the relevant range of parameter space;232

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

Validation of non-linear predictions of -transport and redistribution in unstable regimes,
including potential of coupling to thermal plasma modes either through direct excitation,e.g. generation of seed islands for NTM, or modification of pressure or current profiles;
Tests of active control or mitigation of -driven instabilities.

Studies of NTM physics and application of ECRH/ECCD-based NTM control should have alreadybegun during the non-active and deuterium phases of ITER operation. Control of these modes is
expected to be required to allow achievement of the plasma pressure necessary for the target Q ≥ 10
inductive operation, and thus a relevant demonstration of NTM suppression is required [La Haye,
2006]. However, due to the limitations on available auxiliary power in the D phase it may benecessary for this demonstration to be conducted at less than full (dimensional and nondimensional) parameters. While NTM physics has been extensively studied in smaller devices,
validation of theoretical projections to reactor-scale ρ* is a unique capability of ITER, as is thepossible interaction of NTMs with -driven modes and profile modifications associated with selfheating. The ITER research program into NTM physics during inductive operation in the DT phasewould emphasize physics validation and control optimization in non-dimensional parameter
regimes (β, ν*, ν/ω*, ρ*) not accessible during the D phase, and identify and investigate aspects
associated with -particle effects and self-heating. During hybrid mode operation, ITER research
will investigate the beneficial role of saturated NTMs in maintaining q0>1 at reactor-relevant ρ* andin the presence of -heating.
Non-inductive, high-β plasmas produced in support of the steady-state Q ≥ 5 goal at high q95 arelikely to present more stringent requirements on the NTM control algorithms, as a number ofpotentially unstable modes must be tracked and suppressed as the q-profile evolves. As the kink
β-limit is approached, Δ' is increased, destabilizing the classical tearing mode which can then growas an NTM. To the extent possible, control algorithms will have been tested at similar β andq-profiles, but at reduced absolute parameters, during the non-active and D phases. The extension to
the burning plasma conditions during the active phase of operations will introduce a unique
combination of parameters, as well as confronting coupling of the self-heating to the pressure andcurrent profiles. Optimization of the feedback/feedforward tearing mode control strategy is essential
to realization of the steady-state Q ≥ 5 target. While neither the bootstrap fraction nor the βachievable in ITER will reach the values anticipated for a steady-state Demo based on advanced
scenarios, the interaction of self-heating and -driven instabilities with the NTM control represents
an irreplaceable testbed for validation of models and control algorithms for extrapolation to Demoand reactor conditions.
In the steady-state regime, ITER is expected to operate close to or above the no-wall β-limit andwill therefore require active control of the resistive wall mode (RWM). Much of the underlying
physics associated with control of these modes is known (in principle), and has been, or will be,validated on medium-sized non-burning tokamaks. There will remain some issues, however, which
can only be fully elucidated by experiments at the reactor scale. The first research problem to be
confronted by high-β experiments on ITER is likely to be exploration and validation of models of
the passive (rotational) stability boundary, below which active feedback is not required. Kinetic
effects, including thermal and fast particles, are important elements of the MHD physics of wallstabilization, and of course the rotation in the absence of strong external torque will depend on
transport phenomena. Therefore, experiments at the ITER ρ* and including fast ions (initially from
MeV NBI in the D phase, then by the more isotropic -particle distribution in the DT phase) will
provide essential checks on the theory. A powerful tool supporting such experiments is thecapability to diagnose the proximity to RWM instability by means of active MHD spectroscopy233

ITR-18-003[Chapman, 2009-2]. This could be achieved on ITER by using the in-vessel ELM control coils toapply a rotating n = 1 perturbation while measuring the amplitude and relative phase of the plasma
response and comparing (or fitting) to theoretical models.
The results of these experiments will need to be incorporated into the design of model-based
controller algorithms for feedback stabilization above the wall-stabilized limit [Liu, 2009]. To theextent possible, specific control algorithms and actuators to be used for steady-state operation will
be tested in ITER prior to DT operation by accessing the relevant β at half-field, provided theauxiliary power is adequate in the absence of -heating. Since the extrapolation in ρ* to full field
DT operation will be less than a factor of two, these experiments would provide confidence in theapproach and methodology. The additional RWM physics associated with DT operation will arise
primarily from the addition of an isotropic fast ion () distribution, and from the couplings
associated with self-heating, which imposes additional constraints on pressure and current profiles,
impacting the stability and requiring even more of a MIMO (Multiple-Input Multiple-Output)
control strategy. At a projected target of Q = 5 (Pα = Ploss/2) and fbs ~ 0.5, the ITER experiments
should retain sufficient control authority in the external actuators (possibly including upgrades in
H&CD systems) to sustain the steady-state scenario. Significant experimental time will be required
to validate and optimize the controllers using the installed capabilities, and to explore theboundaries of controllability.
A major focus of the research on ITER should then be to address the more strongly coupled system
anticipated in an ‘advanced scenario Demo’, with higher Q (perhaps approaching 40) and fbs
approaching 0.9, with only weak actuator availability, and operation close to the ideal wall limit.
Aspects of such a research program would include model validation, and could encompass
simulation of strong coupling by feedback on the additional heating and current drive. Upgrades topower supplies of the ELM control coils so that they can provide RWM control to access higher βwould need to be considered to extend the ITER advanced scenario operational regime closer tophysics limits.
A particularly important stability issue for ITER is that of disruptions, due to the potential impact of
disruptions on the research program [Lehnen, 2015-1] (see also section 2.3). In addition to theelectromagnetic and thermal loads imposed on the tokamak internal structures, which can be
estimated to some degree from experience on smaller devices, disruptions at the ITER scale are
predicted to be associated with substantial avalanche runaway electron generation at levels notpreviously encountered. This prediction should be tested already during the PFPO phase on ITER,
together with qualification of mitigation techniques. The technical demands on the mitigation
system during the DT phase are expected to be only quantitatively, not qualitatively, different,
primarily owing to the increased stored thermal energy, except in conditions in which the additionalsources of seed electrons in DT plasmas are sizeable [Martín-Solís, 2017]. Of course, the reliabilityof the mitigation algorithm must be assured before DT operation, and reconfirmed as fullparameters are approached.
While disruption mitigation will be relied upon to prevent melting of the ITER plasma-facing
components and large electromagnetic forces to vessel and in-vessel component, the recovery timeassociated with even mitigated disruptions represents a significant cost to the ITER research
program, especially during the high-Q inductive DT phase. Moreover, while the consequences ofsuch off-normal events in terms of lost research time are serious for ITER, the tolerance for suchoccurrences in a Demo or a reactor will be orders of magnitude less. To this end, research on ITER
should aim to develop real-time control algorithms that avoid (as much as possible) instabilities that
may lead to disruptions and to control methods for stabilizing instabilities that may occur. The
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development of these control techniques and information on the scaling of disruption effects withmachine size will be extremely valuable in the design of the next-generation of fusion devices.
2.6.6.1.4 Multi-scale transport physics
The physics governing magnetic fusion plasmas covers an enormous range of spatial and temporalscales. ITER will represent the first opportunity for the fusion community to assess plasma transport
in a system with a reactor-relevant ratio of the ion gyroradius to the system size (i.e. at very lowρ*). Because ρ* is the only dimensionless parameter governing plasma transport that cannot bematched in present-day devices, the research conducted on ITER in this area will provide
significant leverage in the scaling of present databases to future devices. Another important issue
that ITER will help resolve is the impact of electron transport on overall confinement in reactor-like
conditions in which the -particle energy will primarily be coupled to the plasma electrons.
Detailed turbulence characterization studies on ITER will be more challenging than on present-day
devices. However, some diagnostics seem to be feasible for diagnosing turbulence in ITER (e.g.
Beam Emission Spectroscopy fluctuation measurements) and are being considered for incorporation
into the baseline (see Appendix H). These diagnostics will allow comparisons of plasma turbulencein ITER and those in smaller devices.
Plasma rotation will also be a critical issue for ITER. It is well established that increased plasma
rotation can have beneficial effects on both the transport and stability properties of tokamak
plasmas. Yet, the extent to which ITER plasmas will rotate is still uncertain. Due to the limited
capability to apply external torque with NBI (1MeV energy), self-generated torques may be equally(or more) important than the external torque in generating rotation. Measuring the rotation and,more importantly, identifying the mechanisms responsible for the observed rotation should be ahigh priority research topic on ITER. Such research may point the way to optimization of a future
device to take advantage to the extent possible of the self-generated plasma rotation.
A closely related topic is that of transport barrier formation, sustainment and avoidance. Present
knowledge suggests that these barriers can be formed by a combination of plasma rotation andcompression of the magnetic field on the outboard side of the plasma. While ITER’s tools arelimited in the ability to control rotation, the significant current drive capability of ITER should
allow manipulation of the current drive profile over a wide range to assess the capability to produce
transport barriers in the presence of significant -particle heating. Such knowledge may benecessary to achieve the mission objectives of the long-pulse Q ~ 5 scenarios.
2.6.6.1.5 Physics of the plasma-boundary interface
ITER will represent a major step towards reactor-like edge conditions, simultaneously requiring ahigh pedestal temperature, dissipation of high heat fluxes, low tritium retention, and adequate
exhaust of the helium ash produced by the fusion process. While many aspects of the plasmaboundary interface issue that are required for successful operation of ITER have been demonstrated
individually in current devices, ITER experiments will play a crucial role in obtaining physics datafor developing specifications for power and particle exhaust and plasma-wall interactions in Demo.
Simulations of ITER performance suggest that the strength of the H-mode edge transport barrier
will play a major role in the ability to achieve high-Q operation. At the moment, a first-principles
understanding of the structure of this edge region (i.e. pedestal) is still being developed. The
operating conditions of ITER (low ρ*, low collisionality, high pedestal density relative to theGreenwald density, moderate β, low neutral source) will be distinct from those accessible in current
devices, providing an opportunity to develop more fully the physics understanding of the processes
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that determine the edge plasma structure. Integrally linked to this issue is the control of ELMs to
reduce large transient heat fluxes to the divertor targets. Research on ITER to control the edge
evolution and/or structure through the use of non-axisymmetric magnetic fields (resonant magnetic
perturbations), or other means (for example pellet pacing, vertical kicks, etc.) will take on added
importance as ITER performance increases towards its design goals [Loarte, 2014].
A large part of the Research Plan will be devoted to the study of the performance of plasma-facing
components [Pitts, 2011]. The ITER research program has to demonstrate, to the extent possible,that most of the energy arriving in the divertor region can be radiated, that tritium retention can be
kept within acceptable levels (i.e. below 1%), that adequate helium exhaust can be provided, that
acceptable core impurity concentrations can be maintained, and that plasma-facing materials have atolerable lifetime. In brief, having established an adequate mode of divertor operation that satisfies
the requirements of ITER’s fusion performance mission goals, the ultimate goal of the ITER
Research Plan in this area will be to explore the extent to which this mode of operation can be
developed further to provide a divertor physics and technological solution basis adequate for areactor.
Regarding tritium retention, by the time of DT exploitation, a sufficient knowledge base will havebeen acquired from non-active phase operations so that a robust scheme for fuel retention
management can be developed. Moreover, DT operations will see the deployment and validation of
the global tritium accounting scheme using tritium accounting through the T-plant, determination ofburned fuel through neutron rate measurements, gas balance measurements and local retention
measurements using the planned tritium monitor, which first becomes available in PFPO-2(Appendix H). With the development of long-pulse DT discharges, not only T retention but also
permeation through components to the cooling water can be characterized. While permeation
through beryllium is expected to be negligible, this is not the case for stainless steel components
visible to the plasma (Diagnostic First Wall, TBM front surfaces, blanket module surfaces notcompletely covered by first wall panels, etc.) in which permeation will probably occur during
baking of the in-vessel components and needs to be accounted for to close the tritium cycle.
Although the dpa (displacement per atom) levels in ITER will be relatively modest compared tothose expected in Demo, ITER will provide valuable information on the evolution of the retention
and permeation rate of neutron damaged materials (especially for W and steel) in the tokamak
environment for the next step.
As operations progress towards DT and longer discharges, and as the requirements on disruptivity
get tighter, it is anticipated that during FPO dust production mechanisms evolve from melt
generated droplets to solid particles resulting from the delamination of thick co-deposits. Assuming
a peak Be deposition rate of ~0.2 nms-1 in the divertor (an average of values obtained from
WALLDYN simulations [Schmid, 2015]), a film thickness of ~50 microns will be reached by theend of FPO-1. This is a value for which Be layer delamination has been observed in laboratory
experiments. Monitoring the formation of thick co-deposits and their delamination will be important
to avoid the creation of loose material on surfaces, which could, for example, be detrimental during
plasma initiation.
Regular monitoring of first wall erosion will be important to validate earlier estimates of the firstwall lifetime and inform the need for slight adjustment of the magnetic equilibrium in case of
pronounced localized wall erosion. In essence, it is expected that the wall erosion rate is sufficiently
understood at this stage that the FPO phases will serve only for the validation of existing models,
providing important data for the design of the Demo first wall in terms of erosion lifetime. If, as
expected, long-term wall erosion in Demo devices will be dominated by charge-exchange neutral
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impact, then ITER is likely to be the only device in which this can be (at least partially) verified,largely through the application of models together with the experimentally determined overall
migration balance; the latter requires routine long-pulse operation. Although the experienced gained
through the PFPO phases will be helpful, it is also likely to be only as a result of long-pulse fusion
power operation that proper assessments can be made as to the need and timing of an upgrade to thein-vessel components (see Section 4.4). However, the long timescales for supply of upgraded PFCs
implies that decisions may need to be taken rather early in the nuclear phase.
It is important to note that during FPO-1, the divertor material in the high heat flux region will be
regularly operated close to the tungsten recrystallization temperature. Such conditions are unlikely
to be obtained, even with full power in PFPO-2. Regular execution of a reference pulse (seeAppendix D) will allow the evolution of the divertor power handling capabilities to be carefully
characterized during operations. As the discharge duration is increased, the ion fluence onto thedivertor and the number of mitigated ELMs will also reach very high values. Since the end of
FPO-1 is also likely to correspond to the time where the divertor is replaced, careful analysis of the
exposed material will provide important information on modifications of the material thermal andmechanical properties, supplying guidelines for the possible development of advanced tungstenbased materials for Demo. In general, it is expected that ITER will provide key information on theuse of a divertor with solid targets (as opposed to liquid plasma-facing surfaces) at the reactor scale.
During the FPO-1 phase mixed He-D-T plasma interactions with tungsten at elevated temperatures
will take place for the first time in ITER; under these conditions the formation of ‘fuzz’ is more
likely than during pure He operation in PFPO [De Temmerman, 2018]. The formation of fuzz, orother He-induced morphology changes, might strongly affect the surface emissivity and thus will
require regular calibration of the infrared and visible diagnostics to account for this and to derive
reliable measurements. At the same time, observations of emissivity changes will provide
information on the formation/evolution of any He-induced morphology changes.
2.6.7

Summary of experimental activities and operational time in the FPO Phase

The overall duration of the FPO Phase will be determined by the ITER Members at an appropriate
time in the light of the Project’s success in achieving significant fusion power production, but
considering also the mission goals defined in the ITER Project Specification, the terms of the ITER
Agreement and the Members’ (evolving) strategies towards the exploitation of fusion energy forelectricity production. The research program for the FPO phase developed within the current
version of the ITER Research Plan encompasses only the first 3 DT campaigns with the aim ofpresenting a plausible experimental plan to achieve the principal scientific goals and some elementsof the fusion technology mission, to illustrate the issues and risks that will need to be addressed inestablishing the necessary high fusion gain plasma scenarios and to consider the key aspects ofburning plasma physics to which the ITER program will give access. Clearly, many details of thisplan will evolve in response to the experimental results obtained during ITER operation and in thelight of the experience gained in the study of burning plasmas following the introduction of DT
fuel.
The plan presented in section 2.6 initially emphasizes the development of (predominantly) inductive
scenarios, in particular, the ‘conventional’ ELMy H-mode regime in D and DT plasmas, since this,at present, has the most developed quantitative physics basis for extrapolation to fusion power
production in ITER. The initial D phase of operation is primarily foreseen to allow the developmentof DT-relevant scenarios in parallel with the continuing commissioning of the Tritium Plant and thegradual rise in T throughput, leading to the establishment of plasmas with a sufficient T
concentration to initiate significant fusion power production (say ~ 50 MW). As discussed in the237

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previous sections, much can be learned from these experiments in D plasmas, not least in terms ofscenario optimization, since they should allow easier access (i.e. lower threshold power) to H-mode
operation than the H and He plasmas available during the PFPO phase. Nevertheless, a rapidtransition to DT operation, if allowed by progress in the T-Plant commissioning, would be
favourable for the achievement of high fusion power/ high fusion gain and the associated studies ofburning plasma physics, due to the additional heating power provided by the -particle populationand the further reduction in the H-mode power threshold expected. It should be emphasized that thisemphasis on the conventional ELMy H-mode as the principal path towards the demonstration of
Q = 10 plasmas in ITER will be kept under review in the light of results from the ongoing R&D
programs in current devices and, indeed, in response to the results of initial operation of ITER; it isclear that ITER should exploit the most favourable plasma regime for the achievement of highfusion gain according to the understanding developed at the time of the implementation of the FPO
phase, some 20 years hence.
Progress towards higher fusion power and fusion gain will necessitate the integration of scenario
elements such as fuel mix control, burn control, heat load control, helium exhaust and MHD
stability control, providing capabilities which can be exploited as other burning plasma scenarios
are developed. Several parallel programmatic threads are therefore likely to emerge within the DT
experiments, as approaches to longer pulse operation are developed and specific aspects of burning
plasma physics are explored. In the present version of the ITER Research Plan, it is assumed thatthe important milestone of achieving fusion power levels of ~300 – 500 MW with Q ~ 10, even for‘short’ burn durations of several tens of seconds, will take priority due to its wider implications forfusion energy development. Demonstration of this level of fusion performance would, in allprobability, generate significant support for the longer-term exploitation of ITER and provide
motivation for wide-ranging exploration of ITER’s potential for fusion power production. This
logic forms the basis of the detailed discussion of the early FPO campaigns developed in section
2.6, in which achievement of this initial milestone is followed by lines of research exploring several
modes of long-pulse operation, from conventional ELMy H-mode, through hybrid/improved Hmode scenarios to fully non-inductive operation. The precise evolution of these lines of research
will, of course, depend very strongly on results emerging from initial investigation of such modes ofoperation in the burning plasma regime. The discussion presented here has, therefore, proposed a
strategy based on current expectations, but these expectations will evolve very significantly in the
coming decades. Nevertheless, the achievement of significant fusion power production in ITER will
open the path to an extensive exploration of viable plasma scenarios and to wide-ranging studies ofthe physics of burning plasmas.
In parallel with the scientific program towards high fusion power, high fusion gain, long-pulse
operation and burning plasma studies, the technology R&D program will assume greater
significance in the FPO phase, since this provides a first opportunity to investigate the performanceof numerous Demo-relevant technologies in the burning plasma environment with significant fluxesof 14.1 MeV neutrons (even if the ultimate neutron fluences in ITER will be significantly lower
than in a Demo device or fusion power plant). The progress towards long-pulse operation will
provide insight into the reliability of many components and systems when operating at theirperformance limits over long durations of up to 3000 s. In addition, as presented later in section
3.2.2, the TBM program will make use of these early DT campaigns to test various aspects of TBM
and ancillary system performance, leading to an integrated test of modules capable of breeding
tritium and generating high grade heat, scheduled for FPO-3. This will also be a significant
milestone for the project, being the first practical demonstration of a novel technology, which is
critical for the realization of electricity production from DT fusion.
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Table 2-17 summarizes the activities within research program developed around D and DT plasmas
during the first 3 campaigns of the FPO phase, and provides an estimate of the operational days
required to achieve the objectives of this phase of the Research Plan, in particular to achieve theproject’s mission goal of demonstrating long-pulse (300 – 500 s) operation at Q ≥ 10 and to makesignificant progress towards the mission goal of fully non-inductive operation for durations of up to3000 s with Q = 5. The estimates made for the duration of specific activities, particularly thoserelated to the development of burning plasma scenarios, are necessarily uncertain, since ITER will
furnish the first opportunity for the magnetic confinement program to establish and study thisentirely new regime of fusion plasmas. The challenges which are likely to emerge in thedevelopment of fully non-inductive DT scenarios are particularly uncertain, given the present state
Table 2-17 – Overview of experimental activities and estimates of required operational days
for the FPO research program
Activity
Axisymmetric magnetic control
Plasma scenario commissioning
Kinetic control commissioning
MHD and error field control
Advanced control functions (auxiliary specific, actuator sharing etc.)
Disruption Management
Commissioning of auxiliary systems
Diagnostics
H&CD system commissioning
Divertor and PWI studies
Deuterium plasma studies
Development of D L-mode plasmas to 15 MA/5.3 T
Development of D H-mode plasmas to high power at 7.5 MA/2.65 T
Trace-T H-mode experiments
Initial development of hybrid/ advanced scenarios in D
Optimization of DT H-mode plasmas at 7.5 MA/2.65 T
Optimization of DT fusion performance towards Q = 10 at 15 MA/5.3 T
Extension of DT Q = 10 scenario towards pulse lengths of 300 – 500 s
Development of hybrid/ non-inductive DT plasma scenarios
Control of / current/ pressure profiles in stationary DT plasmas
Optimization of non-inductive current drive at q95 ~ 4 (2.65 T)
Optimization of non-inductive current drive at q95 ~ 4 (5.3 T) - hybrids
Optimization of Q (target Q ~ 5) in long-pulse (~1000 s) hybrids at 5.3 T
Demonstration of 3000 s fully non-inductive (q95 ≥5) plasma with Q ~ 2
Optimization of 3000 s fully non-inductive (q95 ≥5) plasma towards Q ~ 5
Total

239

Days
5
5
28
20
24
11
25
10
15
12.5
104 - 124
32
40
12
20-40
66
180
100
575
25
25
125
100
100
200
1155 - 1175

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of understanding of non-inductive operation in existing devices. As the discussions in section 5.4.5
and Appendix E illustrate, there is much that can be learned during ITER construction and theinitial years of ITER operation, through collaboration with the international fusion community, toestablish a stronger basis for the development of such scenarios. Nevertheless, it is clear from thetime estimates derived for the DT research program that the achievement of these key scientificmission goals is likely to stretch over several FPO campaigns: the estimated duration of the
proposed research program, 1155 – 1175 days, essentially corresponds to the duration of the first 3
FPO campaigns (415 operational days per campaign – Table 2-1). Achieving the mission goal offully non-inductive operation at Q = 5 may also require upgrades to one or more of the ITER
H&CD systems and, possibly, to the power supplies of the ELM control coils for RWM control, inorder to provide both the current drive capability and the flexibility required for the more complex
control environment of fully non-inductive plasmas. Assuming that ITER were successful inmeeting the principal mission goals during the first few DT campaigns (even if there were somedeviation from the precise numerical targets established in the Project Specification) the later FPO
experimental campaigns, beyond the initial campaigns discussed here, will provide opportunities todevelop substantial physics and technology bases for the design and operation of Demo devices (themore successful ITER is, the more likely it will be that a range of Demos will be constructed byindividual ITER Members). Further upgrading of the ITER tokamak and auxiliary systems could,
for example, lead to a focussed program aiming to simulate burning plasma operation under Demorelevant constraints and with Demo-relevant plasma-facing components, H&CD systems,
Diagnostics etc.

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3 Technology Research Program
3.1 Introduction
As outlined in section 1, the ITER project will carry out an extensive research program in fusion
technology in fulfilment of its mission goals and in pursuit of its overall programmatic objective,
defined by the Project Specification: The device should demonstrate the availability and integration of technologies essential for afusion reactor (such as superconducting magnets and remote maintenance); The device should test components for a future reactor (such as systems to exhaust power
and particles from the plasma); The device should test tritium breeding module concepts that would lead in a future reactor
to tritium self-sufficiency, the extraction of high grade heat and electricity production.
Wide-ranging R&D activities have already been undertaken across the entire scope of key fusion
technologies for ITER to provide the technical basis for the design and manufacturing of ITER’s
tokamak, auxiliary and plant systems, e.g. in superconducting materials and magnets, plasma-facing
materials and components, heating and current drive systems, diagnostics, fuelling and vacuum
technology, tritium handling, remote handling, control systems, power supplies, cryogenic andcooling systems, and an ambitious Test Blanket Module Program has been launched to develop andconstruct a series of tritium breeding blanket modules, together with the required ancillary systems,with key features relevant to a fusion reactor.
These technologies will be tested intensively during ITER operations, providing a technical basis
for most of the fusion technology required for construction of a Demo device. The research
program which will be undertaken in each of these technology areas will be developed in the
coming years to ensure that a comprehensive definition is in place in advance of the ITER
Operations Phase. As these plans become available, they will be integrated into the overall
Research Plan in the course of its regular updating. At present, the analysis of the TBM testing
program is considerably more advanced than that of other technology areas, as it is necessary to
specify how the staged testing of various aspects of TBM technology will be integrated into theoperations program (and on what timescale) to provide guidance to the TBM development program;
this must deliver a sequence of TBMs through the operations program, leading to the integrated
testing of tritium breeding and high grade heat extraction during the FPO phase.
The following discussion of the Technology Research Program is therefore limited in the present
version of the Research Plan to the detailing of the TBM testing program, with an initial scope ofabout 10 years, extending from PFPO-2 to the end of the 3rd DT campaign. The longer-term TBM
Program might be expected to undergo significant evolution, as further developments of blanket
concepts emerge in response to the Members’ design studies of potential Demo devices in supportof the exploitation of fusion energy for electricity generation. In addition, it can be expected that thediscussion in this section will be significantly expanded in future versions of the Research Plan toencompass the full scope of technology testing activities which will be implemented during ITER
operations.

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3.2 Test Blanket Module (TBM) Testing Program in ITER
3.2.1

Test Blanket Module Program

3.2.1.1 Background and context for the TBM Program
In order to prepare for Demo, ITER has the mission to test mock-ups of design concepts of tritium
breeding blankets relevant to a reactor, called Test Blanket Modules (TBMs), in three equatorial
ports called ‘ITER Test Ports’. TBMs and associated systems (called Test Blanket Systems, TBSs)
inserted in ITER from the H/He phase are the principal means by which ITER will provide the firstexperimental data on the still open issue on the path to fusion power, the performance of breedingblankets.
The three equatorial ports allocated to TBM testing are the ports # 2, # 16 and # 18. In each of these
ports it is expected to insert a TBM Port Plug (PP) that is formed by a water-cooled steel frame (seeFigure 3.2-1) and 2 TBM-Sets. Each TBM-Set is formed by the TBM and the associated required
shield.
Each TBM must be relevant for the corresponding Demo breeding blanket (in design, materials andoperating parameters). Each TBS includes several associated systems (e.g., primary and secondary
cooling, tritium management and control, and measurement systems). The integration of each TBS
in ITER implies the presence of several interfaces with the ITER machine and buildings that haveto be taken into account in the ITER machine and buildings design [Giancarli 2012].
A large R&D and validation program has to be performed during the ITER construction phase toensure the installation in ITER of six reliable and safe TBSs in order not to jeopardize ITER
performance while producing Demo-relevant testing results.
More than 10 TBS designs have been initially proposed by the ITER Members for testing in ITER.
Among them, six TBSs have been selected for installation in ITER. They will be designed andfabricated by the ITER Members and delivered to the ITER Organization for installation. Each TBS
is developed by a Member Leader (TL) or by a Partnership of Members. The formal way ofcreating a Partnership is discussed in TBM Arrangement (TBMA) Article 12.
The six selected TBSs (including concepts with both solid (SB) and liquid (LB) breeders), their port
allocations, and the corresponding Member Leaders are the following [Giancarli 2016]:
1) Helium-Cooled Lithium-Lead (HCLL) TBS (Port #16/TL: EU). The Helium is operated at
8 MPa and at inlet-outlet temperature of 300-500 °C. The structural material is a ReducedActivation Ferritic/Martensitic (RAF/M) steel, the EUROFER. The Lithium-Lead is a liquid
metal with melting temperature of 235 °C acting as both tritium breeder and neutron
multiplier;2) Helium-Cooled Pebble Beds (HCPB) TBS (Port #16/TL: EU). It makes use of ternary
lithium oxide ceramic pebbles as tritium breeder and Beryllium pebbles as neutron
multiplier. The Helium is operated at 8 MPa and at inlet-outlet temperature of 300-500 °C;
3) Water-Cooled Ceramic Breeder (WCCB) TBS (Port #18/TL: Japan). It makes use of ternary
lithium oxide ceramic pebbles as tritium breeder and Beryllium pebbles as neutron
multiplier. The Water is operated at 15.5 MPa and at inlet-outlet temperature of 280-325 °C;
4) Helium-Cooled Ceramic Breeder Graphite Reflector (HCCR) TBS (Port #18/TL: Korea). It
makes use of ternary lithium oxide ceramic pebbles as tritium breeder and Beryllium
pebbles as neutron multiplier. The Helium is operated at 8 MPa and at inlet-outlet
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temperature of 300-500 °C. A graphite reflector region is added in the back-side of the
TBM;
5) Helium-Cooled Ceramic Breeder (HCCB) TBS (Port #02/TL: China). It makes use of
ternary lithium oxide ceramic pebbles as tritium breeder and Beryllium pebbles as neutron
multiplier. The Helium is operated at 8 MPa and at inlet-outlet temperature of 300-500 °C;
6) Lithium-Lead Ceramic Breeder (LLCB) TBS (Port #02/TL: India). It makes use of 2
coolants: Lithium-Lead (p = 1.2 MPa, inlet/outlet temperature = 300-460 °C) and Helium(p = 8 MPa, inlet/outlet temperature = 300-500 °C). It makes use of 2 breeders: Pb16Li andlithium titanate ceramic pebbles. Pb16Li acts also as neutron multiplier.
The structural material for all the above Test Blanket Modules are Reduced-Activation
Ferritic/Martensitic (RAFM) steels, as these are considered mandatory for a Demo, and for Fusion
Power Plant designs. Note that RAFM steels are ferromagnetic.
The following two additional ‘strategic’ requests were made by the ITER Members and agreed by
the TBM-Programme Committee:
1) The USA requested to reserve the option of leading a TBM concept at a later phase of ITER
operations. Specifically, the USA request noted that nothing in the TBM technical plans
preclude testing of the Dual-Coolant Lithium-Lead (DCLL) concept. To meet this request,
an additional set of inlet and outlet He-coolant connection pipes are designed and installedin Port #18. The associated characteristics would be the following:- Dual-Coolant Lithium-Lead (DCLL) TBS (provisional TL: US). It makes use of 2
coolants: Lithium-Lead and Helium (p = 8 MPa, inlet/outlet temperature = 300-500 °C).
Pb16Li acts also as tritium breeder and neutron multiplier. The structural material is an
RAFM steel;2) The EU requested to not preclude the testing of a Water-Cooled Lithium Lead TBS or of a
DCLL TBS in Port #16. To meet this request, an additional set of inlet and outlet coolant
connection pipes (able to use either pressurized water or Helium) are designed and will be
installed in Port #16. The characteristics of the DCLL-TBS would be very similar to thosedescribed above for the USA; the characteristics of the WCLL-TBS would be the following:- Water-Cooled Lithium-Lead (WCLL) TBS. It makes use of lithium-lead as tritium
breeder and as neutron multiplier. The Water is operated at 15.5 MPa and at inlet-outlet
temperature of 270-325 °C. The structural material is a Reduced-Activation
Ferritic/Martensitic (RAFM) steel, the EUROFER.
ITER may be the only opportunity for testing breeding blankets mock-ups in a real fusion
environment before the construction of a Demo reactor ensuring:i) in the Pre-Fusion-Power Operation: relevant magnetic fields, relevant plasma regimes (e.g.,
H-mode), surface heat fluxes, H (and, possibly, trace D) permeation in the First wall coolant
and disruption-induced loads; this phase will also be essential to learn how to operate a TBS
in a reliable and controlled way from the control room and to confirm several operation datathat will be part of the ITER nuclear licensing file;ii) in the Fusion Power Operation phase: additional relevant neutron flux, volumetric heat, andtritium production with corresponding tritium-management capabilities.
The most serious limitation for blanket testing in ITER is that the magnitudes of neutron flux andvolumetric power density are lower than that expected in a Demo reactor and that ITER operation
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features relatively short pulse length (compared to the quasi-continuous operation expected in a
Demo). The countermeasure is that, for each selected blanket concept, several TBMs have to be
developed making use of ‘engineering scaling’ for testing specific ‘act-alike’ TBMs during
different ITER-phases in order to address different aspects of the TBM performances (neutronics,
thermo-mechanics, thermo-hydraulics, etc.).
In addition, neutron fluence in ITER is too low (up to 3 dpa after 20 years of operation comparedwith about 70 dpa expected in Demo for 2 years of operation) to test long-term irradiation effects onmaterials, and, therefore, they must be determined by other means and in dedicated facilities (e.g.
IFMIF/DONES).
Taking into account these limitations, the major overall testing objectives of the TBM programmein ITER are the following:i) validation of structural integrity theoretical predictions under combined and relevant
thermal, mechanical and electromagnetic loads;ii) validation of tritium breeding rate predictions;iii) validation of tritium recovery process efficiency and T-inventories in blanket materials;iv) validation of thermal predictions for strongly heterogeneous breeding blanket concepts withvolumetric heat sources;v) demonstration of the integral performance of the blanket systems.

Figure 3.2-1: Schematic view of a TBM Port Plug.

TBSs have to be installed early in ITER operation because very important data have to be obtained
during the non-active operational phase (H/He PFPO phase). Within current Integrated Project
Schedule (IPS-2017) most TBMs are expected to be installed during Assembly Phase III in 2031
(see section 3.2.2.1 and Figure 2.2-1: Operations Plan within the Staged Approach.) before PFPO-2.
However for some of them, and for most of the TBSs ancillary systems, the installation may occur
during the Pre-Fusion Power Operation 1 in 2028-2031. The main results expected in the non-active
phase are the following:244

ITR-18-003i) demonstration of the structural integrity of the TBM structures and attachment during
disruptions and Vertical Displacement Events (VDEs);ii) assessment of the impact of RAFM steel, used as the structural material for the most part ofTBMs, on the 3-D tokamak magnetic field structure and evaluation/mitigation of its effects
on plasma confinement, fast ion losses, plasma rotation, etc.;iii) demonstration of all the functionalities of the TBM systems;iv) essential data relevant for safety in order to obtain authorization to proceed with the nuclear
phase of ITER operation.
3.2.1.2 TBM research program accompanying construction
Each installed TBM must be relevant for the corresponding Demo breeding blanket in term ofdesign, materials, manufacturing techniques and operating parameters. In order to ensure theinstallation of reliable TBSs, several specific steps have to be completed in advance. They include
essential R&D activities, relevant for material and manufacturing qualification, safety and licensing
documentation. Moreover, because of the ferromagnetic behaviour of the TBM structures, several
experiments have been and should be further performed in present-day tokamaks to ensure thecompatibility of the TBMs with the foreseen plasma operations and fusion performance objectives
in ITER.
3.2.1.2.1 R&D to be performed on TBMs and associated systems
For each TBS, the main required R&D activities prior to testing in ITER and the corresponding
required time schedule (assuming that the TBSs are installed in Assembly Phase III, around 2031)
are the following: Out-Of-Pile characterization of Demo-relevant structural material and TBM fabrication
process validation (at industrial level); for each TBM, these activities require the selection ofa reference material to be used in all experiments in order to obtain a coherent and complete
set of data. Such a reference material will be used for the manufacturing of all mock-ups andfor the TBMs. To be achieved before the TBM final design reviews planned in the range
2022-25;
 TBS preliminary design; it has to be based on the reference structural material database andit has to be used for deriving the design and manufacturing of the medium/large scale mockups necessary to validate the TBM design. To be achieved in the range 2019-22;
 Fabrication and testing of medium/large size mock-ups of TBMs and TBS components
addressing critical aspects (industrial manufacture or, at least, industrially compatible
manufacture). These tests should be able to guarantee a sufficient reliability of the TBMs
prior installation in ITER. To be achieved by the end of the final design phase; Construction of databases for structural material and joints for use in design codes and codesand standards; these activities correspond to several experimental campaigns able to validate
material properties on a sufficient number of samples for reduction of uncertainties. To beachieved in support of the final design; Construction of databases under irradiation (at least, 1 dpa) for structural material and joints
and for functional materials (e.g., ceramic breeder pebble beds, beryllium pebble beds),
obtained in experimental campaigns performed in fission reactors; these data have to be used
to perform the detailed design of the TBMs. To be achieved in support of the final design;245

ITR-18-003

 Design, fabrication, acceptance tests (expected to be done by the manufacturer) and deliveryof the first TBMs (TBM-EM) to the ITER Organization on the ITER Site expected in 2029
for assembly of the TBM Port Plugs (2 TBM-Sets in each frame), Port Plugs acceptancetests and installation in ITER.
3.2.1.2.2 Effect of TBM ferromagnetic structural material on plasma H-mode performance andfast ion losses
Because of the use of ferromagnetic RAFM steel as TBM structural material, the presence of TBMs
produces localized magnetic field perturbations in the TBM Port area (about 0.6% of the local field
per TBM). Experiments in JET had shown that periodic ripple of the toroidal magnetic field could
have a negative impact on the quality of the plasma confinement in H-mode. Significant
perturbations start at a ripple level of about 0.3%. Since the physics phenomena behind such aneffect have not yet been fully understood, it is uncertain to what extent localized TBM-induced
perturbations have the same impact on the H-mode as toroidally periodic ripple.
In 2010, a Workshop on ‘TBM Impact on ITER Plasma Physics and Potential Countermeasures’was organized by the ITER organization to assess the impact of Test Blanket Modules (TBMs) on
ITER physics operations and to consider possible countermeasures and strategy to be adopted toallow the execution of the TBM Program within the constraints dictated by the achievement of
ITER’s fusion performance goals. Almost 50 scientists participated to the two-and-a-half day
Workshop, including representatives from the ITER Organization, from the TBM teams of all 7
ITER Members, from the international plasma physics community, from the ITER TBM Program
Committee (TBM-PC) and from the ITER Science and Technology Advisory Committee (STAC).
The main conclusions from the physics side were:
1) impact of the expected magnitude of magnetic fields produced by the TBMs is small in low
performance (L-mode) and low-β plasmas; however, the plasma performance may be
degraded in high performance plasmas and, therefore, there is a significant risk of affecting
the first-priority ITER mission of achieving Q = 10;
2) in particular, the experimental results show that ripple causes density pump out andsignificantly reduces plasma rotation, increasing the risk of plasma disruptions;3) theory and modelling of how these magnetic perturbations lead to degradation of plasma
confinement are not yet sufficiently mature and, therefore, the predictive capability islacking to quantify the impact of these issues on ITER high-β plasma performance.
The main conclusions from the TBM design side were the following:1) both the reduction of the TBM ferromagnetic mass and the increase of the TBM recess
distance from the plasma would lead to a significant reduction of the tritium production in
the TBMs that is already at the minimum acceptable level for being able to extrapolate
results to Demo; therefore, the main testing objectives would be jeopardized;
2) use of correction coils in the TBM Port Plugs themselves will add significant further
constraints to the TBM tests (e.g., increase of the replacement frequency due to the limited
coils lifetime under neutron irradiation (less than 2 years)).
The consequences from the TBM-design side have been to fix a recess of 12 cm from the plasma
and a maximum weight of 1.3 tons for the ferromagnetic material present in each TBM.
Recommendation of further decreasing such weight has also been taken as a guideline for the TBM
design development.
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From the plasma-side several experiments were performed in DIII-D, the last occurring in 2014.
The experiments on DIII-D have demonstrated that low-torque ITER baseline scenarios withuncorrected fields introduced by a Test Blanket Module (TBM) mock-up coil are highly susceptible
to rotation collapse and disruption. However, this loss of stability can be recovered (Figure 3.2-2
and Figure 3.2-3) using only n =1 compensation fields generated by ex-vessel control coils similar
to those planned for ITER, enabling sustained operation at ITER performance metrics(NBI ~ 0.7 Nm, Ip/aB = 1.41, βN = 1.8) [Lanctot, 2017]. Companion experiments in high-β plasmas
have also demonstrated that n = 1 error field correction leads to recovery of β and rotation
degradation, and a significant reduction of localized heat loads on plasma facing components
associated with the TBM local magnetic field disturbance. These results validate concepts of asingle dominant mode to the plasma response and show that correction of the n = 1 component
recovers most of the performance impact of the TBM, while optimization of n = 2 fields showed
little further recovery of the plasma rotation. Correction of TBM error fields has enabled full
restoration of low torque access to ITER baseline-like scenarios on DIII-D.

Figure 3.2-2: Evolution of : (a) the TBM coilcurrent, (b) N, and (c) L in low-torque discharges
for three TBM field settings: no TBM field (dottedblue), uncorrected TBM field (dash–dot red), andcompensated TBM field (dashed green and solid
black traces). In every discharge, a 5/4 NTM ispresent when the TBM coils are energized. Onset of
a 4/3 NTM instability in 159811 (green) at 4390 ms
reduces energy and particle confinement, andmodifies the torque balance [Lanctot, 2017].

The DIII-D results provided great encouragement for the viability of the TBM Program on ITER,
and the important knowledge development that it will enable for fusion energy. They show that
TBM error field compensation is essential in ITER, but this can be readily achieved with theexternal n=1 correction coils, recovering performance losses and eliminating associated localizedheat loads. This increases our confidence that ITER will be able to achieve its scientific mission
while developing the TBM technology, which is critical for Demo.
The experimental setting in DIII-D allowed simulating a pair of TBMs in only one location with anequivalent mass to that of the three pairs of TBMs that will be present in ITER. Additional
experiments to take into account the distribution of three pairs of TBMs (in Port #2, Port #18 andPort #16) with TBM mock-up coils in three equatorial ports would be more representative of theactual situation in ITER.

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Figure 3.2-3: Torque access as a function of TBM
field strength expressed in terms of TBM coil current
and total magnetic field ripple for combinations of
TBM and TBM-compensated fields: no n = 1 EFC(red), with optimal n = 1 EFC (green) and withpartial n = 1 EFC (yellow). Region I can be accessed
without encountering an n = 1 locked mode. Regions
II and III are stable only with EFC and inaccessible
otherwise. Region IV cannot be accessed even with
optimized EFC of the TBM field. The dashed line
marks the expected ITER-equivalent torque level in
DIII-D. The total field ripple expected from a pair of
1.3 ton TBMs in ITER together with the TF ripple is
1.2% [Lanctot, 2017].

In any case, it is not likely that experimental tests in present tokamaks and physics/model-based
extrapolation to ITER will provide a fully conclusive evaluation of the effects of ferromagnetic
TBMs on ITER plasma performance, particularly for the high Q scenarios. The operation of the
TBMs during the non-nuclear phase, particularly the comparison of ITER operation without TBMs
in PFPO-1 to that with TBMs in PFPO-2, will probably give the necessary knowledge to understandif the effects of ferromagnetic TBMs on ITER operation and plasma performance can be mitigatedto the level required to achieve the Q = 10 inductive and/or Q = 5 long pulse and steady-state goals
or if more drastic measures need to be adopted for some critical operational phases (e.g.
replacement of the TBMs by non-ferromagnetic dummy TBMs during some operational periods).
Finally, it is important to recall that error fields associated with the use of ferromagnetic RAFM
steels are not an issue in a tokamak where all blanket modules are made of this material, e.g. in a
Demo tokamak, unlike in ITER where RAFM steels are only used for the TBMs.
3.2.2

Experimental program during operation of the TBM Testing Program in ITER

Within the revised ITER schedule based on the Staged Approach, the TBM testing program is
expected to start during PFPO-2, and continue throughout the FPO phase. As discussed in section
2.6.6, the first experimental campaign of the FPO phase will include both D plasmas and the first
DT plasmas, while subsequent FPO campaigns will be developed around a variety of DT plasma
scenarios.
In the operation of the TBSs, the successful operation of the tokamak and accompanying physics
experiments is essential, so that the TBM program needs to be flexible enough to adjust, if
necessary, to the needs of the rest of the ITER program. This includes start-up and shutdown
periods and the adjustment to unscheduled down-times.
The adopted approach for the TBM testing program is to consider it as a ‘piggy back’ experiment
where the TBMs will be exposed to ITER plasma discharges in a range of operational scenarios, aslong as the presence of the TBMs would not impede the goals of the specific physics
experiments/scenario development. This approach implies the need for the TBM teams to workwith the ITER physics/scenario development program closely; this is essential to the success of the
TBM testing program. An example of this kind of requirement is the estimated need of having, inthe last part of an FPO campaign, at least six days of back-to-back pulses in order to reach a248

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sufficient pseudo-equilibrium of the tritium-related parameters in the TBM, thereby being able tomeasure relevant tritium-related data.
3.2.2.1 Overall TBM testing strategy and objectives
The 6 TBSs have to be installed early in ITER operation because very important data have to beobtained during PFPO-2. The majority will be installed with their ancillary systems during theAssembly Phase III, and they are all expected to be completely installed before the end of theAssembly Phase IV shutdown in 2034-35, before the start of the FPO phase (see Figure 2.2-1).
For each of the 6 TBSs types, it is planned to have up to four TBM versions (to be confirmed) to betested in two main phases, a first phase that can be called the ‘learning/qualification’ phase and asecond phase that can be called the ‘Demo-relevant data acquisition’ phase.
The ‘learning/qualification’ phase includes the PFPO-2 and first FPO campaign. It lasts about 5
calendar years and it is mainly devoted to: Verify/qualify the TBSs operation in ITER operational environment. It includes:– checking on how to integrally operate the various ancillary systems such as: i) coolingsystems (CSs) including pressure drops in TBMs and connection pipes, effects on CSs
pipes thermal expansion; ii) liquid metal systems including impact of MHD effects andfilling and draining procedures; iii) detritiation systems (using trace D and/or T); iv)
CODAC systems and connection with the Central Safety System and the Central
Interlock System (for investment protection);– verifying port plug mounting/dismounting operations, Remote Handling systems andrelated operations including in the Hot Cell;– studying effects of magnetic field in all dynamic equipment associated with TBSs (e.g.
pumps, valves, diagnostics, etc.);

Verify the TBS response and the margins for control under abnormal conditions, such asplasma disruptions, Vertical Displacement Events and power excursions. Identify theoperational margins covering all possible operating conditions and confirm/justify them;validation of TBS integrity;

 Solve all ITER/TBS interface issues such as: a) impact of ferromagnetic structures onplasma confinement and ion losses; b) demonstration of the capability of TBMs to withstandplasma disruptions; c) check tolerances with other components; d) check interferences withother ITER systems;
 Collect all relevant data and procedures needed for the licensing process, including
demonstration of the capability to withstand disruption-induced loads. This information isneeded significantly before the beginning of the start of the D-phase in FPO;
 To start preliminary measurements on TBM neutronic responses in order to verify andvalidate neutronic calculations; Demonstrate the coolant capability of the TBMs First Walls;
 Confirm the essential data and validate the assumptions made for the TBMs to be tested in
the nuclear phase (compliance with the ITER license).

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This phase is essentially covered by the operation of the first two TBM versions, the Electro
Magnetic module (EM-TBM), in the PFPO-2 H/He phase and of the second TBM, the ThermalNeutronic module (TN-TBM), in the D-phase and initial DT-phase, i.e. the 1st campaign of FPO.
The ‘Demo-relevant TBS data acquisition’ phase encompasses the later experimental campaigns in
FPO, i.e. after FPO-1 for the purpose of this discussion. It lasts about 5 calendar years and it ismainly devoted to obtaining from the TBSs all data and information needed for the design andmanufacturing of the corresponding Demo breeding blanket. The main expected results in thisphase are the following: Validation of the capability of the neutronic codes and existing nuclear data to predict TBM
nuclear response, including neutron fluxes and spectra, the tritium production rate, nuclear
heat deposition, neutron multiplication and shielding efficiency;
 Investigation of the TBMs thermo-mechanical behaviour at relevant temperatures taking
into account, for the first time, appropriate volume heat sources; assessment of adopted
fabrication technologies, in particular for joints, and the validation of the adopted
manufacturing processes under low-level irradiation; additional information for SB TBMs
on the thermo-mechanics of pebble beds and for LB TBMs on tritium permeation barriers(either coatings or natural oxide layers); Demonstration of the operational behaviour of the blanket components for heat extraction
and for tritium management, including, for instance, the assessment of tritium permeationand of methods for its reduction, of tritium extraction and coolant purification;
 Performance of an integrated test campaign in order to extend the reliability and operational
performance database for the tested breeding blankets in a DT fusion device under Demorelevant operating conditions (except for neutron fluence) for an extended period of time.
This phase is covered by the operation of the third TBM version, the Neutronic-Tritium/ThermoMechanical module (NT/TM-TBM), in the FPO-2 campaign and of the fourth TBM version, theINTegral module (INT-TBM) in the FPO-3 campaign.
It is assumed here to have the same approach for all six selected TBSs. This approach has therefore
to be considered as ‘generic’. In reality, minor differences can be expected from one TBS to anotherto account for their specific characteristics; such variation will be taken into account in a futureversion of the TBM testing plan. The four TBM versions considered for each breeding blanket
concept, to be used throughout the various plasma operational phases, correspond to the four
experimental campaigns shown in Figure 2.2-1 (PFPO-2 through to FPO-3). The detailed objectives
for each campaign are given in the following sections.
In fact, the TBM testing program presented below reflects the adaptation of the initial TBM testing
program, which was defined by the Test Blanket Working Group as an answer to the ITER-STAC-3
issues and presented at the STAC meeting in April 2008. In the test plan presented below, the initial
testing strategy of the four TBM versions has been kept unchanged, although the scope andobjectives of each test campaign have been revised to take into account the loading conditions ineach campaign. Therefore, the TBM Leaders are in the process of further analyzing the impact of
the newly defined ITER operational phases in the Staged Approach and the associated loading
conditions to confirm the number of TBM versions to be tested as well as the associated scope ofthe tests in each phase. The possibility to extend the scope of tests associated with a given version
of TBM or to combine 2 TBM versions into a single one is not precluded at this stage of theassessment.
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Similarly, the tests in the successive campaigns of FPO, after FPO-3, have not yet been clearly
defined. It is, however, expected that they will include the continuation of the operation of the INTTBM systems to address reliability/availability issues and performance of advanced functional
materials (e.g. coatings, beryllides). In this phase it is also envisaged to test other types of TBM
systems, in particular more advanced ones whose technology is not sufficiently mature for being
tested at the beginning of ITER operation.
It is assumed that all four TBM versions will share the same basic architecture, in particular theirstructural part (including the attachment system) whose design will be qualified during the testing
program in laboratory facilities before TBM commissioning and checked/monitored step-by-stepduring the different phases of ITER operation. This strategy ensures a relatively stable interface
between the TBMs and the other ITER tokamak and ancillary systems during the whole operationaltime with benefits for the availability and safety of the machine.
An important difference in the design of each TBM version will concern the integration of thespecific instrumentation and the design of internal components; in particular of the breeding zone,
which could be modified to test optimized design variants or to achieve the required testing
conditions. For instance, the thickness of the breeder material could be increased to achieve Demorelevant temperatures, and the 6Li enrichment could be modified to obtain the desired test
conditions.
Each TBM will include a monitoring system of all the features relevant for the control and safe
operation of the TBM systems. This includes coolant and liquid metal circuit inlet and outlet
temperatures (including by-pass in case of He), pressure measurements in coolant and liquid metal
circuits, mass flow rates and neutronics responses inside the fluids (for the D- and DT operational
phases). The TBM ancillary equipment will also have to be adapted to the different loading
conditions that the various TBMs tested in the different ITER operational phases will be subject to.
3.2.2.2 Testing objectives for the Electro Magnetic Module (EM-TBM)
3.2.2.2.1 Role of the Electro Magnetic Module
Electromagnetic transients in tokamak reactors (e.g. disruptions, vertical displacements) constitute a
severe issue as far as the capacity of the mechanical structure to withstand the induced mechanicalforces and torques is concerned. These electromagnetic loads also play an important role for thegeneral design of the TBMs, almost determining the design and the dimensions of the mechanicalsupports. In past years, several computational procedures and models have been developed andvalidated for the analysis of electro-mechanical transients in tokamak reactors. Some work has been
dedicated to analyze the influence of magnetic materials such as the RAFM steels that areenvisaged for Demo applications. Such steels show a non-linear ferromagnetic behaviour so that the
distribution and magnitude of the loads on the reactor components during electromagnetic transientsare significantly modified in comparison to non-magnetic structures.
Furthermore, ferromagnetic materials also cause distortions of the magnetic field; in ITER the
TBMs magnetized structures produce an increase of the local ripple in the toroidal magnetic field
which can lead to loss of confinement of high energy particles and localized power fluxes on thefirst wall as well as to a possible impact on the quality of the plasma confinement achievable in the
H-mode regime.
Hence, the validation of computational tools suitable for the design of reactor components based onthese materials is an issue. This validation can be carried out to some degree on the basis of suitable
experiments in dedicated laboratory facilities or in existing tokamaks such as ASDEX Upgrade,251

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JET, EAST or KSTAR, etc. However, ITER will offer the possibility of performing tests in whichgeometry, materials, intensity of magnetic fields and plasma disruptions are relevant for reactor
conditions, which cannot be reproduced together in any facility/present tokamak experiments. The
EM-TBM is a mock-up of the Demo blanket having geometry and material composition that are
relevant for EM testing. It will be instrumented so that specific electromagnetic parameters (such as
magnetic fluxes, eddy currents and magnetic forces) will be evaluated from real measurements,
allowing a comparison with calculated values and validation of the models.
For TBMs using liquid metals, an additional field to be studied is the MHD aspects. The liquid
metal flows are affected by MHD effects that can be subdivided into: i) pipe flow in fringing Bfield, ii) flow in manifolds (toroidal and poloidal), and iii) flow between structural components andcooling plates. These flows are sources of several specific issues with respect to pressure drop andvelocity distribution. Electrical coupling between structural components and cooling plates may
cause multi-channel effects leading to additional pressure drops, and also influencing the flow
distribution. For HCLL TBMs, as the liquid metal throughput is quite small, the temperature
gradient in the structural components may generate mixed convection and thus might alter the flow
distribution.
The EM-TBM, seen as the first module of a series of similar modules, will provide a demonstrationof the capability of the structural construction of the TBMs to withstand the mechanical loads
associated with plasma disruptions. As far as heat loading is concerned, relevant tests for the FW
are possible, anticipating the conditions of later phases of ITER operation. Finally, the EM-TBM
will give the possibility to test the overall system and its remote handling (RH) procedures in atritium uncontaminated environment and without activated materials, offering ideal conditions todetect early defects and allow corrections and/or, eventually, repairs.
Therefore, the Electro Magnetic module (EM-TBM) is used in the H/He phase to investigate theresponse of the structure to electromagnetic transients, to test operational functions (start-up andshutdown, cooling systems, remote handling, safety, MHD and trace D-permeation issues for theliquid breeder TBMs (if possible in this phase), etc.) and to collect essential data for the licensing
process to be completed in advance of the transition to nuclear operations. It will also provide
operational experience that would give confidence in being able to operate the TBS with reasonable
reliability during the nuclear phase.
The use of ferromagnetic structural materials will have some effects on the ITER magnetic field
configuration. At present, the most important issue is considered to be the creation of localized
magnetic field perturbations, which may have a negative influence on plasma confinement in Hmode and lead to loss of fast particles and localized power fluxes to the first wall. Since thephenomena associated with these effects are not yet well understood, it has not yet been possible totheoretically assess the potential impact of the EM-TBM in ITER operation and plasma
performance, although experiments in DIII-D are very encouraging regarding the possibility of
mitigating such effects. The present designs for EM-TBMs do not include local active schemes tocorrect such effects (by coils creating a field that nulls the magnetic field perturbation introduced by
the ferromagnetic mass) because it has not been demonstrated that this scheme would be very
effective and because this complicates considerably the design of the TBM Port Plugs. If further
analysis or new experiments would show that it is not possible to mitigate the effects of the TBMs
on the ITER plasma performance to the required level, other possible countermeasures will beassessed (e.g. reducing ferromagnetic material masses in TBMs, further recession of TBMs fromthe plasma separatrix, optimization of the ferromagnetic inserts around the TBM ports).

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3.2.2.2.2 General objectives
 Assess the overall functionality of TBM systems, both in the TBM (e.g., heaters if needed,measurement devices such as thermocouples, sensors, etc.) and in the ancillary equipment
for coolant-circuits (pumps, heat exchangers, flow-meters, thermocouples, etc.) and Liquid
Metal circuit (pumps, valves, tritium extractors, etc.). For the latter, confirm the start-up andshutdown procedures including heat losses, kinetics for heating and cooling of the circuits
and liquid metal draining; Validate the performance and predictability of heat extraction from the first wall, taking into
account the deposited surface heat, thermocouple measurements in structures and the FW
cooling circuit; Validate the structural integrity of the box and of the attachment system (especially during
disruptions and Vertical Displacement Events). This validation is of extreme importance forthe safety dossier and licensing for the acceptability of similar TBMs in the DT phase; Record TBS performance data to be used for confirmation/justification of operational
margins; Assess the effect of reduced activation ferromagnetic (RAFM) steel on the deformation ofmagnetic fields and determination of the induced ions losses with consequent additional heatloads on neighbouring components such as shield blanket modules (expected to be
acceptable from theoretical estimates) and the schemes for their mitigation;
 Assess the impact of reduced activation ferromagnetic (RAFM) steel on plasma confinement
and stability during plasma H-mode operation and the schemes for their mitigation;
 Check and validation of TBM exchange procedures including the RH system both in theport cell and the hot cell, TBM installation/dismantling procedures including TBM
inspection techniques; and possible feedback to relevant procedures and manuals;
 Collect all relevant data and procedures needed for the licensing process.
3.2.2.2.3 Specific objectives/ measurements for the Lithium-Lead TBM Systems
 Measurement of Liquid Metal (LM) pressure drops, velocity profiles and temperature
profiles as a function of LM flow-rate (capacitive pressure transducers, electrical potential
probes, thermocouples, gyrostatic and electromagnetic flow meters) at different flow
channels;
 Measurement of the steady state magnetic field as input for MHD simulations; Measurement of transient magnetic fields and currents in the breeder units to validate
assumptions in EM calculations; Measurement of corrosion products concentration in LM as a function of the LM flow rateand cooling plate average temperature.
3.2.2.2.4 Specific objectives/ measurements for the Ceramic-Breeder TBM Systems
 Measurement of magnetic fields and currents in the Beryllium and ceramic beds to confirm
the design assumptions used in EM calculations.

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3.2.2.2.5 Special requirements for plasma operation scenarios in the PFPO-2 campaign
The EM-TBMs and associated systems are expected to be able to operate in all types of plasma
scenarios achieved in the H/He phase. Few specific requirements from the TBM Program on plasma
operations may be expected for this experimental campaign, ensuring that the data obtained allow
accurate estimates for the worst operating conditions that the TBSs have to withstand in that
operational phase and the following ones. The specific requirements from the TBM Program on
plasma operations cannot yet be fully substantiated for the PFPO-2 experimental campaign. For
example, there should be a consideration whether 25 days of high plasma current L-mode operation
up to 15 MA/5.3 T is a long enough operational time for TLs to confirm essential data and validate
the assumptions made for the design of the TBSs to be tested in the nuclear phase in compliancewith the license requirements, such as safety measures functions expected during a particularincident/accident scenario.
In order to check the heat extraction capability of the FW, it would be necessary to have, for acouple of weeks, plasma pulses with predictable and constant heat flux on the FW. Even if this heatloads are low (<0.1 MWm-2), if they are at fixed value and known in advance, it will be possible toset the appropriate coolant parameters in such a way to have gradients that could be close to thevalue achievable in the nuclear phase. It may well be that specific tests with varying FW surface
temperature would be required for the overall ITER first wall heat load diagnosis.
To have trace-D injected in the later part of PFPO-2 plasma operations would be desirable in order
to measure the D-permeation from the FW (TBMs but also frames, if feasible), and support
assessments of tritium permeation rates during FPO. This would require typically a week or more ofcontinuous operation of the TBMs at defined preheating conditions, and repeated well-definedplasma parameters. There are significant synergies between this type of experiments and thoserequired to check the heat extraction of the TBM FW with those being considered for thecharacterization of fuel retention (see section 2.5.5.12.2.1), so that a common set of experiments
could most likely address both the retention characterization aspects and the EM-TBM specificneeds towards the end of PFPO-2.
3.2.2.3 Testing objectives for the Thermal-Neutronic Module (TN-TBM)
3.2.2.3.1 Role of the Thermal-Neutronic Module
The Thermal-Neutronic module (TN-TBM) is used mainly for investigation of neutronic responses
during the FPO-1 campaign with D-D and initial DT-operation, and to confirm the thermal
behaviour of the TBM FWs. In particular, the TN-TBM main objective is the validation of thecapability of the neutronic codes and existing nuclear data to predict TBM nuclear responses,
including neutron fluxes and spectra, nuclear heat deposition, and shielding efficiency. In relation tothese nuclear responses, this module will permit accurate measurement of tritium production inorder to verify the tritium breeding predictions. It will also give preliminary information on thecapability of the TBSs to extract surface and low-level volumetric heat loads, the latter being
applied for the very first time on the breeding blanket component.
Instrumentation for the TN-TBM will include thermocouples, and activation foil stacks, fission
micro-chambers, and/or scan wires to monitor and assess neutron fluence and energy spectrum.
3.2.2.3.2 General objectives
 Measurement of neutronic fluxes and spectra in selected positions of the TBM. The data will
be used (together with the information derived from other ITER diagnostics) to validate
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neutronic calculations (activation foils, fission chambers and scan wires) and to define
tritium sources; Neutron and gamma load measurements on the FW (bi-chromatic infrared camera +thermocouples, activation foils);
 Assessment of the shielding efficiency of the TBM: neutron flux at the back plate (BP) rearface (e.g. by three neutron detector counters), activation foils in the back-shield layers;
 Calibrate most of the nuclear and thermal measurements equipment (to save testing time
during later DT campaigns);
 Preliminary demonstration of the TBS capability to extract heat loads coming from both
surface heat loads on the first wall and low-level volumetric heat loads in the various blanket
materials;
 Confirm the effects of reduced activation ferromagnetic (RAFM) steel on plasma
confinement and stability during plasma H-mode operation and on induced fast ion losses
and the schemes for their mitigation (as assessed during PFPO-2).
3.2.2.3.3 Specific objectives/ measurements for the Lithium-Lead TBM Systems
 For Pb16Li TBMs, preliminary Pb16Li activation measurements for measurement systems’calibration, including trace tritium measurements;
 Confirmation of tritium-control and management performance (e.g., permeation) using trace
D and/or trace T.
3.2.2.3.4 Specific objectives/ measurements for the Ceramic-Breeder TBM Systems
 Preliminary activation measurements for Be and ceramics;
 Confirmation of tritium control and management performance (e.g. permeation) using D andtrace T.
3.2.2.3.5 Special requirements for plasma operation scenarios in the FPO-1 campaign
The TN-TBMs and associated systems are expected to be able to operate in all types of plasma
scenarios achieved in the FPO-1 campaign. No specific requirements from the TBM Program on
plasma operations are expected for this experimental campaign. The only requirement is to know inadvance the most extreme operating conditions that the TBSs will experience during this phase, inparticular concerning wall loads and allow tuning of neutron flux/fluence measurements in the
TBM, and initial production of Tritium in the breeder zone.
3.2.2.4 Testing objectives for the Neutronic-Tritium/Thermo-Mechanic Module (NT/TMTBM)
3.2.2.4.1 Role of the Neutronic-Tritium/Thermo-Mechanic Module
The Thermo-Mechanic module (NT/TM-TBM) is mainly used for investigation of the TBM
structural behaviour at relevant temperatures and to determine the nuclear responses at Demorelevant 14.1 MeV neutron flux, in particular the tritium production. Typical TBM thermal time
constants are 100 – 200 s for the front part of the TBM and longer for the rear part. Therefore, with~ 400 s pulses thermal equilibrium can be reached, at least in the front part of the TBMs. Typical255

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time constants for tritium control and management are of several days and therefore only
measurements of T-sources are expected to be achieved during this DT operational phase.
Additional objectives are: for the ceramic Breeder TBMs, the thermo-mechanics of pebble beds andtritium extraction by purge gas; for the Lithium-Lead TBMs, investigation on tritium control. These
objectives will be assessed during the FPO-1 campaign, which will have both low and high duty
cycle operation with DT pulses.
Instrumentation for this NT/TM-TBS will include thermocouples, flow-meters, pressuretransducers, stress and strain gauges and T-concentration sensors, and other instrumentation invarious parts of the ancillary circuits and components. In relation to the nuclear responses, theinstrumentation will include activation foil stacks to monitor neutron fluence and energy, fission
micro-chambers, and/or scan wires and measurement systems allowing the determination of Tritium
inventories.
For the Lithium-Lead NT/TM-TBMs the main objectives are twofold: First is the measurement of
the thermo-mechanical characteristics of the TBM (temperature and stress/strain distribution) under
neutron flux to validate calculation assumptions and results. Second is the measurement of the inlet
and outlet tritium concentration (both in the coolant and LM circuits) that will be carried out inorder to validate a local model simulating the tritium transfer in the TBM and allowing thedetermination of a global tritium mass transfer coefficient. This global mass transfer coefficient will
characterize the global permeation from the LM to the He circuits through the RAFM steel structure
in a system model. Furthermore, thermal measurements could be used to estimate the actual heat
flux on the first wall and the nuclear heat deposition. Then, an equivalent Neutron Wall Loading(NWL) could be estimated and the tritium production could be calculated. This procedure would
allow checking if tritium production could be assessed using temperature measurements. However,for all measurements related to tritium the interpretation of the obtained values is quite complex due
to the pulsed operation with a quite long exhaust time.
For the Ceramic Breeder TBMs, the NT/TM-TBM will be mainly dedicated to the testing of the
Pebble bed thermo-mechanics, a critical issue for the temperature control of this kind of concept
and to ensure correct performance of the T purging flow. The thermo-mechanical behaviour ofpebble beds is determined by numerous effects, e.g. the differential thermal expansion between thebeds and the structure, elastic and plastic deformation of the pebble bed and structure, thermal
creep, ‘bed ratcheting’, gap formation, and irradiation swelling. These effects play an important role
for the design-relevant properties of the beds, in particular the thermal conductivity, the bed-to-wallheat transfer coefficients, and the modulus of elasticity. The design of a blanket using a breeder andmultiplier in form of a pebble bed requires the development of efficient computer codes. The
validation of these codes and of the implemented models is an important part of the development
work which has to be carried out on the basis of suitable experiments, up to now either out-of-pileor in fission reactors. However, these experiments suffer from severe physical and technicalrestrictions, especially the lack of the non-uniform volumetric heating. Hence, tests underconditions closer to fusion power reactor blankets are necessary and the test in ITER will inherently
provide such conditions. Of particular importance for thermo-mechanical pebble bed testing are theneutron flux and spectrum in large volumes, and the related volumetric heat sources in the DT
phase, which allows the installation of test modules with pebble beds of representative size. A
limitation of the experimental validation can be the length of the burn pulses; preliminary
calculations have shown that a full steady state condition in the overall TBM is reached only forpulses of about 1000 s; however, some TBM regions have a shorter time constant (~10 s in the FW
vs. the surface heating flux) so that the use of shorter pulses can be used for this kind of validation.
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In any case, the main restriction of this kind of experiments in ITER is the relatively low neutron
fluence which implies that end-of-life conditions (in comparison to Demo) in the pebble beds
cannot be reached.
Other appropriate test conditions must be established by selecting appropriate design parameters.
One important parameter is the maximum temperature in the pebble beds. The relatively low
neutron flux and resulting power density can be compensated by increasing the thickness of thepebble beds or increasing 6Li enrichment (in comparison to a Demo design). First estimates haveshown that the thickness of the breeder pebble bed in ITER must be about 25 mm at 90% 6Li
enrichment to achieve the same maximum breeder temperature (~920°C) as in a power reactor
blanket. Another design parameter is the orientation of the pebble beds with respect to gravity; incurrent Demo blanket concepts, the beds are horizontal at the tokamak mid-plane with theinclination increasing in the poloidal direction. In single-null divertor reactors, vertical beds exist at
the top of the poloidal cross section of the torus (opposite the divertor x-point). In vertical beds,
gravitational forces additionally affect the thermo-mechanical behaviour of the pebble beds, andratcheting effects may lead to bed compaction and void formation. The experimental investigationof this phenomenon should be included in the thermo-mechanics test program.
3.2.2.4.2 General objectives
 Measurement of temperature and stress/strain distributions in selected parts of the TBM
structure in order to validate computational codes and modelling; Assessment of the thermal stresses on joints and corresponding joints’ behaviour to validate
computational codes and modelling; Further validation of the performance and predictability of heat extraction from the firstwall, taking into account the deposited surface heat and thermocouple measurements instructures and in the FW cooling circuit; Validation of the performance and predictability of heat extraction from the structures andbreeding zone, taking into account the deposited volume heat (for the first time);
 Measurement of tritium and gamma production (activation foils, fission chambers) forcomparison with the neutronic calculations in order to assess the tritium breeding capabilityof each blanket design; Assessment of the tritium permeation towards the coolant and initial validation of tritiumextraction techniques; Irradiation of samples of structural and functional material for determination of material
nuclear responses. Measurements of Neutron Fluxes (> 1 MeV), α and proton production
rates near the FW and welds (activation foils, fission chambers) for material damage
estimations.
3.2.2.4.3 Specific objectives/ measurements for the Lithium-Lead TBM Systems
 Influence of the Pb16Li flow rate and chemistry control on Pb16Li and He average tritium
concentrations; For Pb16Li TBMs, Pb16Li activation measurements (T with gamma spectrometer in line
and integral, Po and other reaction products with counters and spectrometers);
 Influence of the 6Li enrichment and, therefore, of the volume ratio of 6Li/RAFM steel.
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3.2.2.4.4 Specific objectives/ measurements for the Ceramic-Breeder TBM Systems
 Test of the thermo-mechanical behaviour of the pebble beds under different configurations.
Measurement of the temperature field and (if possible) stresses. Comparison withcomputational code predictions; Tests of fluid-dynamic performance of the purge flow, including measurement of pressure
drops and investigation of possible causes of performance degradation during operation due
to pebble fragmentation and deposition of dust;
 Test of the effect of the purge gas flow rate and purge gas chemistry (hydrogen addition,
impurities) on tritium extraction from the breeder material.
3.2.2.4.5 Special requirements for plasma operation scenarios in the FPO-2 campaign
The NT/TM-TBMs and associated systems are expected to be able to operate in all types of plasma
scenarios achieved in the FPO-2 campaign. The first requirement for the TBMs is to know inadvance the most extreme operating conditions that the TBSs will experience during thisoperational phase. A further important requirement in the FPO-2 campaign is to have several seriesof pulses with a good in advance knowledge of their characteristics in order to adapt the TBM
operating parameters to the expected loads and durations.
3.2.2.5 Testing objectives for the INTegral TBM (INT-TBM)
3.2.2.5.1 Role of the INTegral Module
The INTegral TBM (INT-TBM) is used to demonstrate the operational behaviour of the blanket
components for heat extraction and tritium management during the FPO-3 experimental campaign(DT operation at high duty cycle). The overall objective of the integrated test campaign is to extend
the reliability and operational performance data base for the tested breeding blankets in a DT fusion
device under Demo-relevant operating conditions (except for neutron fluence) for an extended
period of time.
Instrumentation will include thermocouples and flow-meters, pressure transducers, stress and strain
gauges, tritium concentration sensors. In this case, the installed instrumentation will be reduced, in
a compatible way with the required measurements, in order to reduce the impact of the detectors on
the TBM functionality.
The INT-TBM should perform its designated mission without failure, enabling parametric studies
related to thermal response, prove thermal and mechanical predictions, demonstrate the on-line
tritium handling capability and provide general experience for system improvements.
The INT-TBM ‘looks-like’ the corresponding Demo blanket module, suitably scaled in order to fit
in the Test Blanket port. It represents, to the largest extent possible, all synergetic effects acting in ablanket module of a power reactor, such as e.g., cooling the FW and other structures, local heatgeneration and transport and tritium generation and release. The low Neutron Wall Load (NWL) in
ITER compared to Demo necessitates compromises, where necessary, between the look-alike Demodesign in favour of an act-alike Demo design (e.g. multi-pass helium coolant channels in the FW) in
order to achieve reactor-relevant operating conditions. These are achieved mainly in terms oftemperature level and distribution (both in the box and in the breeder zone), coolant conditions suchas inlet/outlet temperature and pressure or coolant parameters such as Reynolds and Nusselt
numbers and, for the Pb16Li-based TBM, Pb16Li relevant parameters, such as Hartmann, Stuart
and Sherwood numbers.
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This campaign will cover testing of the tritium systems (on-line extraction and removal) in a long
period of long pulse operation and back-to-back series of pulses (about six days of continuous backto-back pulses are required). Pulses of several thousand seconds would be required to approachsteady state conditions for the thermal performance of the TBM. The objective is to validate adynamic model of the tritium transfer in the TBM system and obtain relevant data allowing an
extrapolation to a Demo blanket concept and its technological components.
3.2.2.5.2 General objectives
 Validation of the on-line tritium processing systems designed for the different TBM
concepts. This will require a relatively long period of back-to-back series of reference pulses(about six days of continuous back-to-back pulses) as well as the development of on-line
tritium measurement techniques/sensors with suitable accuracy;
 Validation of a dynamic model of the tritium transfer in the TBM system through relevant
experimental data allowing an extrapolation to Demo (blanket concept and technologicalcomponents);
 Parametric analysis on the impact of the main circuits’ functional parameters (purge gas
flow-rate, He chemistry, efficiency of tritium permeation barriers) on the global tritium
system performance in order to obtain relevant information for extrapolation to Demo.
3.2.2.5.3 Specific objectives/ measurements for the Lithium-Lead TBM Systems
 As a first step, the model of the global mass transfer coefficient of the TBM will be finalized
and validated. Tritium concentration inlets (in He and Pb16Li circuits) will be considered as
inputs. Tritium concentration outputs will be measured and compared with calculations; As a second step, the predictive capacity of the tritium system model and component
recovery models will be assessed. Only system functional parameters and tritium recovery
parameters will be the considered as inputs, the overall inlet and outlet concentrations ofboth He and Pb16Li concentrations will be compared with calculations.
3.2.2.5.4 Specific objectives/ measurements for the Ceramic-Breeder TBM Systems
 Monitoring of temperatures and stresses in the solid breeder zone and of the pressure drop in
the purge systems;
 Test of the effect of the purge gas flow rate and purge gas chemistry (hydrogen addition) on
tritium extraction from the breeder.
3.2.2.5.5 Special requirements for plasma operation scenarios in the FPO-3 campaign
As noted above, in this experimental campaign the TBM testing program will require about six daysof continuous back-to-back pulses. In this period, particularly towards the end of FPO-3, ITER longpulse plasma operation should be well developed so that disruptions will be infrequent. Ideally,long pulses at high fusion power and with a period of 6 days (to be confirmed) of back-to-backpulses, e.g. with high FW heat loads, should be dedicated to TBM testing. Pulses of several
thousand seconds would be required to approach steady state condition for the thermal performanceof the TBM.

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3.2.2.6 Overall achievements in the TBM Testing Program from PFPO-2 to the FPO-3
campaign – implications for later operations
As illustrated in Figure 2.2-1, the revised schedule within the Staged Approach foresees (wherestart and end dates are inclusive): First Plasma: December 2025 – June 2026;
 PFPO-1: December 2028 – May 2030;
 PFPO-2: June 2032 – February 2034;
 FPO campaign 1 (DD + first DT): starts December 2035 (16 months campaign);
 FPO campaign 2: starts December 2037 (16 months campaign);
 FPO campaign 3: starts December 2039 (16 months campaign).
The schedule outlined above assumes that the long-standing pattern of ITER operational campaigns
developed in the original RAMI analysis will be maintained.
The transition to ‘partial’ nuclear operation at ITER begins with the introduction of tritium into the
Tritium Plant, which is foreseen in present planning to coincide with the start of Assembly Phase
IV, and an initial regulatory authorization is required for this step. Authorization for full nuclear
operation of the ITER facility is expected later, towards the end of Assembly Phase IV.
The majority of TBSs are expected to be installed during the Assembly Phase III in 2030 - 2031 andthe planned operations for the four TBM-versions are expected to last until March 2041. From thepoint of view of the TBM Program, these 10 calendar years can be divided in two large phases, afirst phase that can be called the ‘learning’ phase and a second phase that can be called the ‘Demorelevant data acquisition’ phase.
The ‘learning’ phase includes the PFPO-2 and the FPO-1 campaign 1. It lasts about 6 calendar years
and it is mainly devoted to: Verify/qualify the TBSs operation in the ITER operational environment. It includes:– checking how to operate the various ancillary systems such as: i) cooling systems
including pressure drops in TBMs and connection pipes, effects on CS pipe thermal
expansion; ii) liquid metal systems including impact of MHD effects and filling anddraining procedures; iii) tritium-detritiation systems (using D); iv) CODAC systems andconnection with the Central Safety System and the Central Interlock System;– verify port plugs mounting/dismounting operations, RH handling systems and related
operations including in the Hot Cell;– effects of magnetic field in all dynamic equipment associated with TBSs (e.g. pumps,
valves, diagnostics, etc.); Identify the operational margins covering all possible operating conditions andconfirm/justify them; validation of TBMs integrity;
 Solve all ITER/TBS interface issues such as: a) impact of ferromagnetic structures onplasma confinement and fast ion losses; b) demonstration of the capability of TBMs towithstand plasma disruptions; c) check tolerances with other components; d) checkinterferences with other ITER systems;260

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 Collect all relevant data and procedures needed for the licensing process, including
demonstration of the capability to withstand disruption-induced loads. This information isneeded significantly before the end of the PFPO phase; To start preliminary measurements on TBM neutronic responses in order to verify andvalidate neutronic calculations; Demonstrate the coolant capability of the TBMs First Walls.
This phase is essentially covered by the operation of first TBMs, the Electro Magnetic module(EM-TBM), in the PFPO-2 campaign and of the second TBM, the Thermal-Neutronic module (TNTBM), in the FPO-1 campaign.
The ‘Demo-relevant data acquisition’ phase concerns the FPO-2 and FPO-3 campaigns involving
experiments in DT plasmas. This phase lasts about 4 calendar years and is mainly devoted toobtaining all data and information from the TBSs relevant for the design and manufacturing of thecorresponding Demo breeding blanket. The main objectives of this phase are the following: Validation of the capability of the neutronic codes and existing nuclear data to predict TBM
nuclear response, including neutron fluxes and spectra, the tritium production rate, nuclear
heat deposition, neutron multiplication and shielding efficiencies;
 Investigation of the TBMs’ thermo-mechanical behaviour at relevant temperatures taking
into account, for the first time, appropriate volume heat sources; assessment of adopted
fabrication technologies, in particular for joints, and the validation of the adopted
manufacturing processes under low-level irradiation; additional information for SB TBMs
on the thermo-mechanics of pebble beds and for LB TBMs on tritium permeation barriers(either coatings or natural oxide layers); Demonstration of the operational behaviour of the blanket component for heat extraction
and for tritium management including, for instance, the assessment of tritium permeationand of methods for its reduction, of tritium extraction and of coolant purification;
 Performing an integrated test campaign in order to extend the reliability and operational
performance database for the tested breeding blankets in a DT fusion device under Demorelevant operating conditions (except for neutron fluence) for an extended period of time.
This phase is covered by the operation of the third set of TBMs, the Neutronic-Tritium/ThermoMechanical module (NT/TM-TBM), in the FPO-2 campaign and of the fourth set of TBMs, theINTegral module (INT-TBM) in the FPO-3 campaign.
These data and information will allow the design and manufacture of a Demo breeding blanket,
provided that high neutron fluence material data are obtained in parallel in another facility; since theexpected total neutron fluence in ITER is much smaller than the one expected in Demo, the neutron
fluence-induced effects on materials cannot be assessed in ITER.
As a consequence, the objectives for later FPO campaigns (i.e. FPO campaigns after FPO-3) can be
twofold depending on the fusion program strategy assumed by each Member:

Those Members supporting a ‘fast track’ approach to fusion energy development will likely
proceed immediately to the detailed design and construction of Demo using the information
on TBSs described above. In this case, the later FPO campaigns can be used to continue thetests on the same TBSs as in the earlier DT campaigns in order to collect further data useful
to determine the reliability of the corresponding breeding blanket system;261

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

3.2.3

Those Members having a longer-term approach to fusion energy development may wish totest more advanced breeding blanket concepts in ITER in order develop high-performance
breeding blankets. In this case, the FPO campaigns after FPO-3 will be partially used torepeat the operations performed in the previous phases for the initial TBSs but now for the
new advanced concepts. The constraints are that the new advanced TBSs should use thealready existing pipes, connections, and equipment.
Summary and Conclusions

In this preliminary adaptation of the TBM Program to the revised ITER schedule within the Staged
Approach, the previous TBM testing strategy of testing in sequence four TBM versions for eachtype of TBM designs has been maintained.
It must be stressed that this proposal may be modified in the future when the following uncertainties
have been resolved:
1) Impact of the TBM ferromagnetic structural material on plasma performance - if nodefinitive experimental evidence is available before testing of the TBMs in ITER and/or if
the effects of TBMs are identified by comparison of PFPO-1 and PFPO-2 plasmas but
cannot be satisfactorily mitigated, it may be necessary to add a partial use of dummy TBMs
made of non-ferromagnetic materials during the PFPO-2 phase until mitigation schemes forthe effects of ferromagnetic TBMs have been developed and implemented;
2) Eventual duration of the different ITER TBM phases - the possibility of extending the scopeof tests associated with a given version of TBM, or to combine 2 TBM versions into a single
version with a single testing campaign, cannot be precluded at this stage of analyses;
3) Delivery date of each Test Blanket System – It may happen that some TBSs are delivered
later than the others.
Nevertheless, it can be concluded that the integration of the TBM Program into the Staged
Approach schedule is possible, confirming its robustness for adaptation to modified ITER
operational plans without jeopardizing the TBM testing objectives.

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4 Upgrade Options and Required R&D
The aim of the proposed upgrade plan within the context of the ITER Research Plan is twofold:(i) to minimize the risks to the achievement of ITER’s goals by implementing timely upgrades ofsystems in view of the results obtained from the scientific exploitation of ITER itself and by the
R&D carried out by the ITER Members during construction and the initial phase of ITER operation;and (ii) to expand the capabilities of the ITER device by suitable upgrades of ancillary systems
emphasizing the demonstration of physics and technologies that will be required in Demo.
The specific upgrades which could be envisaged at present and their rationale are provided in thefollowing sub-sections. The corresponding time-windows within which upgrade decisions for key
systems contributing to the achievement of ITER’s goals could be taken, based on preceding
operational experience, are described within the deliverables for sections 2.5.4, 2.5.5, 2.6.3, 2.6.4
and 2.6.5; these are summarized below. It should be noted that the implementation of some
upgrades requires significant timescales (~5 years), as described in Appendix G for the H&CD
systems. Early decisions may be required, therefore, to avoid delays in the progress of theexperimental program during DT operation, when such upgrades may well be essential to thesatisfactory completion of the ITER mission. Table 4-1, at the end of this section, provides an
overview of the main options for upgrades in each of the major categories of tokamak and auxiliary
systems which can be foreseen at present as potentially required (or simply ‘desirable’ for improved
physics studies) at each stage of the IRP.
Possible required upgrades following the results of the PFPO-1 operational campaign:
PFPO-1 will provide first operational experience for: L-mode plasmas up to 7.5 MA and for
H-mode plasmas up to 5 MA with ELM control by the application of 3-D magnetic field
perturbations. On this basis the following upgrades could be identified at this stage:


Need to increase the additional heating level: if the power required for H-mode operation
exceeds expectations – there might be a preference for ECRH or ICRF (if the corresponding
antenna is available), depending on operational experience;

Need to upgrade the pellet injection system: to provide ELM control capabilities, if ELM
control by 3-D fields is found inefficient (for stationary H-modes or for transient H-mode
phases, i.e. access/exit and with varying q95);

Possible need to upgrade the NBI system with the 3rd HNB: if ELM control with no-torqueinput (only RF heated plasma discharge with ECRH and, possibly, ICRF will be performedin this phase) proves difficult or leads to excessive confinement degradation.

Possible required upgrades following the results of the PFPO-2 operational campaign:
PFPO-2 will provide first operational experience for: L-mode plasmas in the range 8 - 15 MA,
H-mode plasmas around 7.5 MA with ELM control by the application of 3-D magnetic field
perturbations, and current drive capabilities for long-pulse operation in ITER. On this basis thefollowing upgrades could be identified at this stage (if the assessment of PFPO-1 is not conclusive):

Need to increase the additional heating level: if the power required for H-mode operation
exceeds expectations, with the preference among ECRH, ICRF or NBI options being
developed based on operational experience;263

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

Need to upgrade the pellet injection system: to provide fuelling and ELM control
capabilities, if either fuelling is found insufficient at higher plasma currents or ELM control
by 3-D fields is found inefficient (for stationary H-modes or for transient H-mode phases,
i.e. access to/exit from the H-mode and with varying q95);



Need to upgrade the NBI system with the 3rd HNB: either for ELM control (if ELM control
in H-modes with 2 HNBs proves difficult, or excessive confinement degradation is observed
due to the low input torque), or to achieve the required current drive for steady-state
operation (if, for any reason, the current drive provided by the HNBs is lower than
expected);Need to upgrade the ECRH system: if the level of current profile control in the current rampup and flat-top is insufficient for long-pulse/steady-state scenarios in DT, or therequirements for NTM control by ECCD make power sharing demands on ECRH difficult
to manage at the baseline power level (i.e. the baseline ECRH power level provides
insufficient scenario flexibility when using ECRH both for core heating and NTM control).

As detailed in Appendix G, a decision on the installation of the 3rd HNB upgrade taken at this stage
would lead to a significant delay of the FPO phase, whereas a decision at the end of PFPO-1 is notexpected to have any impact on the start of FPO.
Possible required upgrades following the results of the first FPO operational campaign:
FPO will provide first operational experience for D and DT plasmas, in particular for high-Q
operation at 15 MA, and the first experience on the issues which must be resolved for thedevelopment of the Q ~ 5 long-pulse and steady-state scenarios. By this time most of the upgrades
required to achieve these goals should have been implemented. However, a need for further
upgrades might be identified at this late stage from the exploration of high- DT plasma conditions,which can lead to requirements that cannot be met by the installed systems at that time and thatwould not have been identified by previous operation in PFPO-1/PFPO2:
Need to upgrade the pellet injection system: to provide DT fuelling mix control and ELM
control capabilities, if either DT fuelling or mix control with the (baseline) 4 pellet injectors
is found insufficient in DT plasmas, or ELM control by 3-D fields is found inefficient (forstationary H-modes or for transient H-mode phases, i.e. access/exit and with varying q95);



Need to upgrade the ECRH system: if the level of current profile control is insufficient forthe development of long-pulse/steady-state scenarios in DT, or if the required power level toprovide W accumulation control and NTM control in high-Q plasmas exceeds the baseline
power level of 20 MW;



Need to upgrade the ICRF system: if studies on the effect of heating mixes reveals that ahigher level of ICRF power optimizes Q in integrated H-mode scenarios;

Need to upgrade power supplies for Resistive Wall Mode control: if the capability provided
by the baseline power supplies for the ELM control coils is not sufficient to ensure therequired level of RWM for high-Q/high- scenarios.

In addition to these upgrades, others may be required to ensure robust ITER operation, i.e.
essentially disruption-free, but with routinely successful mitigation when disruptions do occur. The
need for these upgrades (vertical stability control, disruption mitigation upgrades, etc.) is likely tobe identified as the expansion of the ITER operational space proceeds towards: higher plasma
currents (PFPO-2) and plasma energies (PFPO-2 and FPO), and with the start of DT operation
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ITR-18-003(FPO). However, without ITER’s operational experience, it is not possible to provide a specificpoint in the experimental program where the need for such upgrades will be identified.

4.1

Heating and current drive systems

As specified in the Project Requirements [ITER_D_27ZRW8, 2014], The maximum installed
power foreseen for H&CD in ITER is 130 MW, of which a maximum of 110 MW can be injected
simultaneously into the tokamak due to limits in the total site power for steady operation. A review
of analyses of the exploitation of H&CD systems in candidate non-inductive plasma scenarios in
ITER is given in Appendix E, while options for upgrading of the H&CD systems are discussed inAppendix G. The Project Requirements identifies options for 4 H&CD upgrade scenarios, leading
to a total installed power of ~130 MW:
1. The addition of 20 MW each of ECRH, ICRF and LHCD to the baseline capability;
2. An upgrade of NBI to 50 MW with the addition of a 3rd HNB line, plus the addition of
20 MW of ECRH and LHCD; the removal of one ICRF antenna while upgrading of theperformance of the other to maintain 20 MW of IC;
3. An upgrade of NBI to 50 MW with the addition of a 3rd HNB line, plus the addition of
20 MW of ECRH and ICRF;
4. An upgrade of NBI to 50 MW with the addition of a 3rd HNB line, plus the addition of
40 MW of LHCD; the removal of one ICRF antenna while upgrading of the performance ofthe other to maintain 20 MW of ICRF.
As discussed in Appendix G, allocation of ports to meet H&CD and Diagnostics requirements
within the Staged Approach has resulted in the deprioritization of LHCD as an H&CD upgrade
option. Priority in the development of H&CD upgrades is now given to:

A 3rd HNB line to increase the NBI power from 33 MW to 50 MW;



An increase of the ECRH power from 20 to 40 MW (maintaining the same number of ports);

An increase of the ICRF power by bringing the two antennas from 20 to 40 MW.

For the baseline heating systems, the estimated time required to perform the upgrades of existing
H&CD systems (from procurement to installation) is in the range of 5-6 years. Therefore, a decision
immediately after PFPO-1 is required in order to have upgraded systems ready by the beginning of
the FPO phase. The choice of upgrade options would then depend on experimental observations andoperational experience during PFPO-1, as well as on R&D results from ITER Members' facilities.
These experimental observations are expected to provide the first quantitative information regarding
the required level of additional heating to achieve ITER’s Q ≥ 10 goal. In addition, uncertainties
regarding the required conditions for H-mode access and stationary type-I ELMy H-mode operationin ITER, in particular its dependence on isotope and plasma species, are expected to be reduced by
R&D studies. Proposals are already under development to ensure that H&CD upgrades could be
implemented at a sufficiently early stage to allow their exploitation early in FPO, as discussed inAppendix G.
4.1.1

Baseline H-mode access capabilities

Although the technical feasibility of delivering 10 MW of ICRF during PFPO-1 has not yet been
demonstrated, in the context of the IRP it is assumed that the 10 MW of ICRF power will beavailable for PFPO-1, and therefore that 30 MW and 73 MW of auxiliary heating power should be265

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routinely available for the PFPO-1 and PFPO-2 campaigns respectively. Access to H-mode in
helium (and possibly hydrogen) is expected during PFPO-1 within the program based on the
5 MA/1.8 T option once the installed H&CD systems are commissioned for short pulse (a few tensof seconds) operation. However, stationary type-I ELMy H-mode operation in helium and hydrogenplasmas at higher currents and fields may require the full baseline auxiliary heating power foreseen
in PFPO-2.
In the case of hydrogen plasmas, the lack of a reliable heating scheme for ICRF at 7.5 MA/2.65 T is
an issue and, thus, under these conditions the maximum level of coupled power from this system
may be significantly lower than that specified (10 MW in PFPO-1, 20 MW in PFPO-2). Therefore,
during PFPO-2, operation in helium at half field is expected to provide the best experimental
evaluation of the auxiliary H&CD power required for Q = 10 operation in ITER. This information
would complement observations from 5 MA/1.8T hydrogen and helium plasmas during PFPO-1
and PFPO-2.
4.1.2

Strategy towards H&CD upgrades

A decision on upgrades to the existing heating systems to ensure the success of the Q ≥ 10 mission
could be based on experimental results from the PFPO-1 phase. The objective of such as decision
would be to have the upgraded systems ready for DT operation to achieve the Q ≥ 10 goal as earlyas possible. If the results from experiments conducted in PFPO-1 were unclear, any decision on
H&CD upgrades would be postponed until PFPO-2, in which case the upgraded heating systems
would only be available later in the FPO campaigns, possibly influencing the timescale for theachievement of the Q ≥ 10 goal. However, delaying a decision on the preferred H&CD upgrades
until PFPO-2 could also cause a delay in the transition to the FPO phase, if further optimization ofplasma scenarios were required in advance of DT operation or if installation of upgraded (for some)H&CD systems were to be undertaken prior to activation of the tokamak structures – the strategy
for upgrading of the HNB system discussed in Appendix G is designed to avoid this latter issue.
Even if early operational experience confirms that the baseline H&CD power will provide anadequate basis for meeting the Q ≥ 10 mission, it would still be sensible to plan for an H&CD
upgrade, with priority given to improving the current drive capabilities of the device to achieve theQ ≥ 5 steady-state goal, once the initial inductive DT experiments reach the Q ≥ 10 milestones. The
choice of preferred current drive upgrade (NBCD, ECCD) will depend on: (i) the results of researchin ITER on the real capabilities of each H&CD system, in particular the level of the current drive
obtained with the existing systems (HNB and ECRH); (ii) the experimental evaluation of the role
played by plasma effects on the localization of the driven current in ITER compared with standard
resistive current diffusion; and (iii) R&D carried out at the ITER Members’ facilities on thedevelopment of plasma scenarios for steady-state operation in ITER, as described below.
4.1.3

R&D needs to support decisions on H&CD upgrades

In parallel with the analysis of H&CD upgrade proposals, several R&D activities would need to be
pursued to provide an adequate basis for meeting the Q ≥ 10 baseline and Q ~ 5 steady-state
operation goals.
1. Identification of the EC current drive efficiency and requirement for an ECRH system
upgrade (to 40 MW) to provide more power and greater flexibility for other applications, forexample, active control of NTMs, which will become a potential issue during the highperformance H-mode operation in PFPO-2; depending on the progress of commissioning
and operation of the ECRH plant in advance of First Plasma, a staged upgrade of the ECRH
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system could start earlier (after First Plasma), which would allow additional power for
PFPO-2 and FPO. In particular, increased power for off-axis current drive for NTM control,
while maintaining 20 MW for central heating, would significantly enhance the flexibility for
scenario development and control;
2. Identification of the ICRF power coupling to the plasma during PFPO experiments, and therequirement for an ICRF system upgrade (to 40 MW) for handling issues related to fast ion
driven modes, such as Alfvén Eigenmodes;
3. Identification of the NB current drive efficiency and influence on fast ion driven modes, andthe requirement for a NBI system upgrade (3rd HNB) to provide more power and current
drive for long-pulse and non-inductive steady-state operation;
4. Optimization of H&CD mixes for high performance, non-inductive, steady-state operation
within plasma stability boundaries.

4.2 Fuelling and pumping upgrades
SOLPS modelling for the reference 15 MA ITER scenario indicates that the required pumping
speed for acceptable helium concentration (an edge concentration of 3%) and power load at thedivertor is typically <60 m3s-1. This pumping speed is about 40% lower than that foreseen for 300 500 s Q ≥ 10 pulses, for the neutral pressures in the range of several Pa expected in these
conditions. Thus, an upgrade of the foreseen pumping capability is not expected to be required for
He removal, at least for Q = 10 operation. Despite this, and because our understanding of neutral H,
D, T and He behaviour under ITER divertor conditions is not very accurate, it will be important toperform helium pumping investigations early in the non-active and deuterium campaigns (inH-mode for helium and deuterium) so that the currently predicted performance can be confirmed.
A further issue which could lead to a reconsideration of the pumping requirements in ITER is thatof density control. For the long pulses which ITER targets, wall saturation is expected, so that the
only effective particle removal will be provided by active pumping. The baseline requirement for a
maximum throughput of 200 Pam3s-1 (~1023 atom·s-1) is significantly larger than the expected coreplasma outflux on the basis of global scalings for Q = 10 plasma conditions (Nplasma/E < 31022 s-1),
and is, in fact, similar to that in present devices. For example, ITER-like discharges on JET with<ne>/nGW ~ 0.85 and Ip = 2.5 MA have Nplasma = 71021 particles with E = 0.4 s and a steady
fuelling/ pumping rate of ~ 21022 s-1. The main difference between present devices and ITER isthat the recycling particle flux is typically a factor of ~10 larger than the core particle outflux (equal
to the pumped flux for stationary conditions) in current tokamaks, while it is a factor of 100 for
ITER Q = 10 plasma conditions. Thus, particle control in ITER is subject to larger uncertainties
with respect to recycling flux behaviour, such as outfluxes from saturated walls, than present
devices. Early operation in the non-active phase of ITER, particularly in helium (and in H-mode),
for which pumping is lower than in hydrogen, will show whether or not there is a need to increase
the initial pumping speed in order to achieve satisfactory plasma density control in the expectedconditions for Q = 10.
An increase of a factor of 1.5 in pumping speed can be easily achieved by modifying the reference
operation scheme of the cryopumps. The ITER pumping system operates with 6 isolatable
cryopumps connected to pumping ports at the divertor level. For Q = 10 operation, the system isoperated with 4 pumps active during the discharge while the other 2 are being regenerated andbrought back to operation for the next plasma. For high neutral pressure operation, in which thesystem is not conductance limited, the pumping speed can therefore be increased by operating all
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pumps simultaneously. An additional increase of a factor of 2 (i.e. a factor of 3 in total with respect
to the baseline, if all pumps are used simultaneously) is possible, if the pulse duration is reduced this would be compatible with the theoretical maximum value of the peak fuelling rate, though theoverall consistency with the operation of the fuel reprocessing cycle would need to be confirmed.
The drawbacks of this high pumping speed operation mode are that: (i) the discharge duration
would need to be shortened to meet safety requirements (in this respect, the deflagration limit in thepumps is likely to be more important than the total in-vessel tritium inventory limit) and to maintainthe total throughput per discharge within the capability of the Tritium Plant; (ii) the discharge
repetition rate would be affected, since all pumps would have to be regenerated between pulses.
If the pumping speed needed upgrading beyond this factor of 3 higher than baseline level, thiswould have considerable implications for ITER operations (possibly including further shortening of
the pulses to remain within the limit on the time-averaged throughput of the Tritium Plant). The
practical implementation of such an upgrade would need to be investigated in detail, but currently
the most viable route would rely on improvements in cryopump design and/or technology and thereplacement of the initial cryopumps by others with higher pumping speeds. Otherwise, additional
ports would need to be allocated to pumping, which requires a major redistribution of theequipment and applications foreseen for the lower ports in ITER – the feasibility of such areconfiguration of the port allocation, together with the required access through the cryostat, would
need to be assessed in detail, in particular for the period following the initiation of DT operation.
Implementation of the pellet fuelling system at ITER will be carried out in a phased approach, sothat upgrading of the system is foreseen and requires a relatively simple implementation procedure.
The initial pellet injection system for PFPO-1 will have two injectors that can individually injecthydrogenic (H in PFPO and later H/D/DT in FPO) pellets at a frequency of 4 - 16 Hz with a pellet
volume in the range 17 - 92 (±20%) mm3, covering the needs from pellet pacing (of ELMs) to pellet
fuelling with two injection geometries (low and high field sides). Within the Staged Approach, thenumber of pellet injectors is planned to increase by 2 before PFPO-2, to ensure reliable fuelling of
ITER hydrogenic plasmas. The pellet injection capability is further upgradeable to a system of up to
6 injectors, which can be used simultaneously for both pellet pacing and fuelling. Initial
experimental results of the capabilities of the system will be obtained by hydrogen pellet fuelling ofhydrogen plasmas and helium plasmas (in L-mode and, as far as possible, in H-mode) during the
PFPO phase of ITER operation, although for helium H-modes, the hydrogenic content needs to be
kept low. Tests of pellet pacing of ELMs at high injection frequencies of up to 32 Hz (in PFPO-1)
and 64 Hz (in PFPO-2) will be performed for helium and, if possible, hydrogen type-I ELMy Hmode plasmas. The expected frequency of uncontrolled ELMs in 7.5 MA/2.65 T DT plasmas is
3 Hz [Loarte, 2014]. He H-mode plasmas with similar densities and ratios of Pin/PLH to those in D
plasmas have a similar ELM frequency [McDonald, 2004; McDonald, 2010]. Therefore, it isexpected that for He plasmas at this level of plasma current the uncontrolled ELM frequency willremain at around several Hz. These tests will most likely be performed in PFPO-2, where a pellet
injection frequency of up to 64 Hz can be achieved.
Depending on the outcome regarding pellet requirements for pellet pacing during the PFPO phases
and the associated effective fuelling with hydrogen of helium plasmas, experiments on pellet pacingin helium plasmas may be possible in stationary conditions or only for timescales of several
seconds, the timescale on which the concentration of H in the He plasmas is expected to evolve. In
either case, the results obtained in helium and, where possible, hydrogen plasmas, together with theresults of ELM control using 3-D fields in stationary H-modes and in transient access/exit phases,
will provide a basis for decision-making on the choices for the upgrade path to be followed that will
be confirmed by further experiments in deuterium.
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Upgrade of the pellet fuelling and ELM pacing system to its full performance (6 injectors) can be
implemented on the timescale of a few years, so that a timely decision during PFPO-2 would allow
the system to be operational for DT operation, or possibly even for the D phase. This upgrade willallow simultaneous use of pellets for pacing at frequencies in the range of ~40 Hz (as predicted tobe required for producing controlled ELMs of an acceptable amplitude) together with fuelling(frequencies of 4 - 8 Hz [Loarte, 2014]). As part of this upgrade, the maximum injected pellet size
could be increased, if required, for more effective fuelling; this could allow ITER to explore
regimes at 15 MA with densities at or beyond the Greenwald density.
A more substantial upgrade of the pellet system would be required if the baseline LFS pellet
injection geometry/ size/ velocity prove inadequate for ELM control by pellet pacing, either
because ELMs are not triggered by these pellets, or because the pellets required to trigger ELMs inthis geometry lead to excessive plasma fuelling. In this case, a more extensive use of the HFS
injection geometry for pellet-pacing, or the installation of a LFS high speed injector of small pellets
could probably need to be considered for ELM control.
The fuelling upgrades described above can lead to operation of ITER beyond the maximum averagedesign throughput of 200 Pam-3s-1 for some experimental conditions. It is foreseen that the durationof flat-top phases of the pulses in such experiments would be shortened (from the reference 300 500 s) as required to ensure that the in-vessel inventory of all hydrogenic species condensed on the
cryopumps during a plasma pulse, as well as the in-vessel tritium inventory, remains within safe
limits. The average total gas throughput per pulse must also be kept within the limits foreseen forthe operation of the Tritium Plant.

4.3 Diagnostic upgrades beyond the 2016 Baseline
The possible diagnostic upgrades beyond the 2016 baseline for ITER can be grouped into several
categories.
Firstly, one can consider upgrades of the specifications of present systems, for example to increase
their time/space resolution, or their radiation hardness, or the completeness of the system in a moreor less straightforward fashion, in particular for these cases:







High Temperature Core Thomson Measurements;
Edge Thomson Scattering coverage;
Fast pellet tracking through fast Visible/IR cameras;
Development of rad hard soft X-ray detectors, neutron detectors, preamps, opticaltransmission, ADC;
Equipment for laboratories in the Hot Cell facilities;
Contingency for systems which do not meet performance requirements.

Secondly, ‘new’ diagnostics whose usefulness for ITER is acknowledged, but for which complete
funding is not yet available or for which satisfactory design solutions are still under development,
can be considered. As discussed in Appendix H, some of these systems, particularly those related tofusion products, will already be ‘enabled’, for example via the installation of in-vessel components:






Fibre Optic Current Sensor (FOCS);
Collective Thomson Scattering;
Radial and Vertical Gamma Ray Spectrometer;
High Resolution Neutron Spectrometer and Tangential Neutron Spectrometer;
Measurements of lost -particles;269

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




Measurements of dust and retained tritium;
In-Vessel Transporter (IVT) or Multi-Purpose Deployer (MPD) related tools, e.g. forneutron calibration and MPD tools for PWI investigations;
IR bolometer system to increase the diversity of that system;
Water leak detection spectroscopic wide angle viewing system.

In the third category are upgrades of diagnostic capability that are perhaps less straightforward, butwould enhance the device performance, for example, by progressing from a simple measurement toa measurement that can be used to improve tokamak control capability, or from inputs into theplasma control system (PCS) to inputs for investment protection systems in a more direct way suchas:Vis/IR and bolometers, planned to feed into PCS, could be upgraded to feed more directly
into investment protection systems.

The fourth category includes those new diagnostics installed for demonstration in fusion-relevant
conditions and intended for eventual implementation in Demo, or to demonstrate operation of ITER
in a Demo-like mode such as:Radiation hard imaging solutions in Visible and IR range;
Bolometers which are substantially ‘harder’ against radiation;
New diagnostics for helium concentration and Ti measurements which are not based on
CXRS;
Diagnostics for steady-state operation (new magnetic sensors, etc.).

The upgrades to installed diagnostics are likely to be stimulated as a consequence of plasma
operation as the ITER program develops and plasma measurements become available. These are
expected to consist of hardware improvements (improved signal-to-noise ratio, detection limits,etc.) and/or improvements in space/time resolution. Most ITER diagnostics are already foreseen to
be upgradeable in this way and forward planning in this respect can be carried out on a relatively
short timescale (~1-2 years) if no port plug modification is needed for the upgrade. This would
increase to 5 years if a new port plug were required.
An additional driver for upgrades will be the continued development of the control aspects ofvarious measurements from the operational experience on ITER. Related to this are the expected
incremental improvements to the CODAC interface. This development of control aspects isimportant as it will help to determine the ultimate utility of the various diagnostics for tokamak
operation. As just one example, the primary diagnostics for q-profile control (at present) in ITER
are MSE, polarimetry, and the underlying equilibrium reconstructions based on the magnetics
ensemble. How well these measurements work in concert will determine their utility for control, as
opposed to a simple determination of the q-profile post hoc. A corollary to this is the value of earlyexperience –the deployment of much of the diagnostic ensemble should be undertaken as early as is
feasible from the schedule and budget perspectives. Similarly, delays in the commissioning of thefull NBI capability will delay the ability to make MSE measurements. While not critical in the early
operational phases of ITER, such measurements will play a role in determining the q-profile control
capability for the later phases.
On a longer timescale, development of new diagnostics towards Demo should be considered. This
involves tests of diagnostics for helium concentration and Ti measurements which are not based on
CXRS, diagnostics for steady-state operation (new magnetic sensors, etc.), as well as possible
modifications to present diagnostics for their use in control schemes aimed at Demo control
demonstration, such as increased resolution of soft X-ray cameras, polarimetry, etc. For the higher
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Paux regimes (~110 MW) that one could foresee for Demo-like operation simulations, the fast particle pressure is expected to increase by a factor of at least 1.4 with respect to that foreseen in thereference scenarios. It is possible that such pressure increase may have a noticeable influence on the
MHD stability of the discharges, as the reference scenarios are expected to be close to stability
thresholds in this respect. With this outlook, upgrades to diagnostics of -particles and, inparticular, the fast -particle population, are an obvious option. This could imply two types ofactions: (i) increasing the lines of sight and/or power launched into the plasma for collective
Thomson scattering measurements of the perpendicular -particle energy distribution, and/or (ii)installing a new viewing geometry that allows direct diagnosis of the parallel -particle energy
distribution. This second option would most likely require the installation of the necessary
waveguides before the active phase of ITER exploitation.

4.4 Upgrades of plasma-facing components and materials
Regarding PFCs, upgrades naturally separate into modifications to the divertor and first wall. The
modular divertor design and the possibility for in-situ replacement of the First Wall Panels (FWPs)
protecting each blanket module mean that upgrades can proceed independently and in all cases fully
by remote handling [Merola, 2014].
4.4.1

Divertor

Before the decision to eliminate the first CFC/tungsten divertor in 2013 [Pitts, 2013], the original
upgrade path for the ITER divertor foresaw two further full divertor exchanges, with the first,
a full-W variant installed in the pre-nuclear shutdown and a third unit (reusing some components of
the non-activated first divertor set) planned for deployment after the first ~10 years of nuclear
operation. Two divertor replacements are still retained in the revised schedule in the case of ITER
extended operations (see [ITER_D_3T6DZ9, 2015]), but no decision has been taken as to thedetails of future upgrades. What is clear is that design-to-manufacturing and assembly completion
requires a minimum of ~7 years, and so this time interval must be respected for any new divertor
upgrade. In the current baseline planning, design activities for the second divertor would begin at
the start of the PFPO-1 campaign.
The present full-W divertor, to be installed from the start of PFPO-1 operations, will be qualified
according to defined numbers of load cycles to 10 MWm-2 (~5000) and 20 MWm-2 (~300, qualified
as ‘slow transients’), themselves derived from earlier versions of the Research Plan in which
operation phases were broken down into H-He I, H-He II, DD and DT [Carpentier, 2014]. The
revised plasma operation in the present version of the ITER Research Plan, consistent with theStaged Approach, anticipates that these cycle numbers could be attained within the first ~2 years of
the FPO campaigns (section 2.6.4) and, thus, according to the present definition of divertor
engineering-lifetime, this would be the logical point to replace the first units. It may also be
coincident with the time at which other upgrades (e.g. H&CD systems) are being installed.
Given the long development and manufacturing lead times, it is clear that a divertor upgrade in amatter of years following the start of FPO could not benefit from much operational experience
regarding exposure in the tokamak environment. Moreover, early non-active operation cannot attainhigh heat fluxes, significant transient energy densities or the largest electromagnetic forces.
Nevertheless, it is expected that the lessons learned from the manufacture, assembly and installationof the first divertor will contribute enormously to the development of the second, even before thecomponent is exposed to plasma. If a decision is taken to replace the divertor after only a relatively

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short period of nuclear operation, it is therefore inevitable that the second divertor will resemble thefirst preserving, in particular, the vertical target geometry.
A significant research effort is currently devoted to the development of new tungsten-basedmaterials, aiming, for example, at improving the material ductility, reducing the possibility for
macro-crack formation and propagation [Coenen, 2016; Pintsuk, 2016], and mitigating theformation of plasma-induced morphology changes [El-Atwani, 2014]. This research is linked
primarily to expected Demo requirements, with the tacit assumption that ITER will continue using
pure tungsten as a plasma-facing material. Given the wealth of research in this area, and the factthat the recent development of high flux/high fluence plasma generators allows the testing ofmaterials/components to extremely high fluences, it is important to keep the option open for the useof an advanced tungsten-based material for any future divertor upgrade in ITER.
Based on current materials R&D and issues already experienced with regard to the design andprototyping of the first divertor, the following options could be considered:






Splitting the inner and outer targets into separately cooled high heat flux and baffle regions;such an option was already proposed qualitatively in the 2011 Deferral Workshop at thetime that the option to eliminate the CFC/W variant was being considered (see Physics basis
for the first ITER divertor [ITER_D_6JCJ5T, 2011]);
Use of more advanced W materials for the monoblocks in the high heat flux areas;
Modifications to monoblock surface shaping based on early experience in the PFPO
campaigns;
Upgrades/refinements/replacement of the first set of in-vessel diagnostics installed ondivertor cassettes.

Regarding a third divertor, in addition to the further refinement of materials which will doubtless be
possible as R&D proceeds through the ITER construction and first operational years, the major
modifications are most likely to be in the area of divertor geometry; one potentially important issue,
however, is radiation damage at the CuCrZr cooling pipes for a divertor exposed to higher neutron
fluences – to some extent this may already be problematic for the second divertor.
Two main geometrical changes can be foreseen within the interface constraints of the present
cassette structure. The first concerns modification of the vertical target and/or dome geometry to
optimize operation at high fusion power and acceptable helium exhaust. The second would be acomplete change to a horizontal plate geometry, maximizing plasma volume and plasma current
capability within the design limits of the PF system. Such a modification would also likely requirechanges to the blanket module geometry in the lower outer regions of the poloidal cross-section. An
upgrade of this type is only likely to be performed if experiments in the first years of FPO show thatradiative power dissipation in burning plasma conditions is effective and is only weakly dependent
on divertor target geometry and/or that the SOL power e-folding length in ITER (and therefore that
foreseen for Demo devices beyond ITER) is substantially larger than presently foreseen. If this doesturn out to be true, ITER operation will have demonstrated that the problem of power handling isless severe than currently foreseen for thermonuclear plasmas. In such a case solutions which
optimize the volume available inside the vacuum vessel for use by the plasma should be examinedin ITER.
4.4.2

First wall

The FWP design is such that replacement of the entire first wall material is possible once in thedevice lifetime. From an engineering qualification point of view, the Be wall panel lifetime is272

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designed on the basis of 50% of the full 30,000 cycle budget for the central solenoid full flux swing.
An obvious long-term upgrade, in view of support to a conventional Demo, would be the switch toa full high-Z wall (either tungsten or steel), allowing ITER to perform Demo-like plasma-wallinteraction studies (albeit at low wall temperature), and to examine the compatibility of a high-Zwall at the reactor scale in terms, for example, of edge power load control studies in the absence of
low-Z intrinsic impurities and core high-Z impurity control. Such studies are naturally linked toheating system upgrades allowing ITER to approach Demo plasma conditions more closely.
An upgrade to a full-W wall may also profit from the advances in materials described above andfrom the ongoing assessments of the most suitable Demo first wall material. Developments in
power handling technology may also be such that specific panels could be deployed in high heat
flux areas, notably in the upper main chamber, opening up the possibility of quasi-double null
operation and possible access to small ELM regimes, should ELM control prove to be insufficienton ITER in baseline configurations. The first upgrades to power handling technology and materials
may, in fact, occur in a gradual approach, whereby the transition to full-W begins, for example, inthe midplane (start-up) and upper main chamber regions. Such upgrades may even be envisaged at
the end of the PFPO phases, for installation during the pre-nuclear shutdown. This might benecessary if damage occurs on certain Be panels (for example due to melting in the current quench
phases of disruptions), or if early operation demonstrates that higher heat handling capacity isrequired. This will only be possible, however, if a decision is taken early enough to allow
production of the required hardware. It should be mentioned that the Enhanced Heat Flux first wall
panels use a hypervapotron cooling technology which is also used for the divertor reflector plates
and dome, while the Normal Heat Flux Panel use a steel pipe liner embedded into a CuCrZr heatsink. Although the divertor reflector plates and the dome umbrella use hypervapotron technology
with flat tungsten tiles, dedicated R&D will be needed to develop a tungsten FWP, depending on
the required thickness and size of the individual tungsten tiles.
Depending on the observed rates of T permeation through steel during ITER operations,modification of steel surfaces visible to the plasma might be required. Several options could beconsidered, such as coating of the existing steel surface with a permeation barrier material,
complete replacement of the 316L(N), etc. All will require a dedicated development/qualification
program to ensure they are compatible with operation conditions. Such an upgrade could be
performed during the replacement of the FWP.

4.5 Upgrades of disruption mitigation systems
The present ITER strategy for disruption mitigation is based on the injection of large quantities ofhigh-Z atoms to dissipate the thermal and magnetic energy through radiation. Whereas a high levelof confidence exists that this scheme can mitigate the electromagnetic and the thermal loads, there
is a significant uncertainty as to whether it will prevent runaway electron formation or will be able
to mitigate the consequences of a runaway beam once it forms. The main technique chosen for the
ITER disruption mitigation system (DMS) to inject the impurities, and possibly also large quantities
of hydrogen or deuterium for runaway avoidance, is the Shattered Pellet Injection (SPI). The ITERDMS design has the maximum flexibility in terms of quantities, species and injection location thatis possible within the technical constraints. However, there is the risk that the mitigation
performance is insufficient at some point during the execution of the research program whenapproaching high current or high energy operation. In that case, an alternative scheme or a possible
upgrade option must be available to minimize the impact on the execution of the experimental
program. Therefore, it is essential that in parallel to the implementation of the DMS baseline
design, R&D studies are performed that aim at exploring alternative injection or mitigation
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concepts and which prepare the efficient use of the ITER DMS. Besides alternatives to the present
baseline DMS, an upgrade could also involve the use of different injection locations or an increasein the injection capabilities.
Alternative injection concepts would aim at improving the deposition of the injected material inside
the plasma, on increasing the assimilation efficiency and on allowing flexibility in the injected
species. Possible candidates, such as compound pellets, shell pellets or rail gun systems, would
require further exploration to allow their performance to be characterized and to come to a moremature design stage. Alternative mitigation schemes would aim at techniques other than material
injection. These could include, for example, the use of wave-particle interactions or passive
protection schemes.
Any of the above mentioned upgrades of the DMS will result in major impact on the operationalschedule and on other systems which are integrated in the relevant port plugs. Minor upgrades, suchas changing the pellet sizes of the existing injectors, are possible without significant impact on theoperational schedule.

4.6 Upgrades to plasma magnetic control (verstical stabilization, shape, error
fields, resistive wall modes)
4.6.1

Plasma vertical stabilization

Reliable plasma vertical stabilization (VS) is essential to limit the frequency of VDEs and, as aconsequence, to reduce the impact of thermal and electromagnetic VDE loads on the machine
lifetime. The capability of the VS system also determines the operational space for plasma
axisymmetric magnetic control, in the terms of maximum plasma elongation and maximum value ofthe plasma-wall gap on the plasma outboard side for given plasma current and plasma conditions. A
capability to increase this gap may be important for first wall protection, taking into accountpossible static and dynamic errors in axisymmetric magnetic control and 3-D distortion of theplasma boundary caused by various sources of non-axisymmetric magnetic fields (e.g. currents inthe ELM control coils). For example, according to the study reported in [Chapman, 2014], theRMP-induced 3-D distortion of the plasma boundary foreseen in the ITER 9 MA steady state
scenario can be up to ±5 cm.
The level of low frequency noise in the measurement signal of the vertical velocity of the plasma
current centre, dz/dt, used in the feedback loop for stabilization of plasma vertical displacements, isan important parameter affecting plasma vertical stabilization [Gribov, 2015]. In ITER, the relevant
frequency band lies below ~300 Hz. A high level of the noise in the dz/dt signal: (i) increases theprobability of VDEs during operation with low values of the stability margin, ms, (e.g. operation
with high plasma elongation and/or with a high value of the plasma-wall gap on the plasma
outboard side); (ii) increases heating of the VS in-vessel coils and their busbars; (iii) causes
disturbances in plasma-wall gap control (due to coupling of the ‘fast’ plasma vertical stabilizationand the ‘slow’ gap control), increasing the amplitude of oscillations in the plasma-wall gap control(the dynamic error in gap control); (iv) increases AC losses in the superconducting magnets, due tothe increase in amplitude of current and magnetic field oscillations in the superconductors, andincreases joule heating in the structures surrounding the superconductors. If necessary, measures for
reduction of the low frequency noise in the dz/dt diagnostic signal should be taken based on ITER
operational experience.
In nominal operation, plasma vertical stabilization is provided by the VS in-vessel coils (the verticalstabilization system VS3) in combination with the system VS1 (which controls the differential274

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current in the external coils PF2, PF3 and PF4, PF5). In this mode of operation, VS1 is used for
reduction of the time-averaged component of the current in VS3 [Gribov, 2015]. If necessary (e.g.
for deliberate plasma vertical oscillations generated with the aim of ELM triggering in low current
H-mode plasmas, [Gribov, 2015]), an upgrade of the vertical stabilization system VS3 could be
performed, increasing the velocity of the cooling water in the VS coils and the on-load voltage ofthe VS3 power supply (a modification of the VS3 power supply is required).
Algorithms for in-service assessment of the fatigue lifetime of the VS in-vessel coils, based on the
waveforms of currents flowing in the coils, should be developed. The residual fatigue lifetime of thein-vessel VS coils should be periodically reassessed based on the data obtained in the course of
ITER operation. It will also be important to gain experience of plasma vertical stabilization without
the in-vessel coils, but using only a stabilizing feedback loop based on VS1, should the residual
fatigue lifetime of the VS3 coils assessed on the basis of operating experience be predicted at somepoint to be insufficient for the entire ITER operational life, or in the case of failure of the VS3 coils.
Experimental optimization of the plasma magnetic configuration (in particular, the maximum valueof the separatrix elongation and location of the strike points), allowing operation without the VS3
coils while retaining a low probability of VDEs, as well as optimization of the VS1 stabilization
algorithms and of the dz/dt diagnostic, should be performed, preferably at low plasma currents,
where the impact of VDEs on the machine is significantly lower.
If necessary (e.g. when the residual fatigue lifetime of the VS3 coils becomes low, or when the coilsfail), the plasma stabilizing system VS1 can be upgraded. This upgrade consists of an increase in
the VS1 power supply on-load voltage from 6 kV to 9 kV, which allows an increase of the plasma
vertical stabilization operational boundary when using only the VS1 ex-vessel coils by about 25%[Gribov, 2015]. Another upgrade for VS control involves an additional system utilizing ex-vessel
coils to further improve plasma vertical stabilization. This system, referred to as VS2, varies thedifferential current in the coils CS2U and CS2L. The VS2, with an on-load power supply voltage of
6 kV, allows an additional increase of the operational boundary (relative to that for VS1 with avoltage limit of 6 kV) by about 25% [Gribov, 2015].
Depending on the need, these upgrades of the vertical stabilization capability could be implemented
as early as in the shutdown period between PFPO-1 and PFPO-2.
4.6.2

Control of plasma current position and shape

Should it turn out that the voltage requested from certain coil power supplies for the control ofplasma current and plasma-wall gaps in key ITER plasma scenarios can potentially exceed theinstalled voltage capability, the number of converters in the corresponding circuits can be increased(within the voltage limits for the coil insulation tests). Depending on need, this could be considered
as early as in the shutdown period between PFPO-1 and PFPO-2.
4.6.3

Control of error fields

18 ‘saddle’ superconducting correction coils with 9 independent power supplies will be used forminimization of error fields with toroidal mode number n = 1. In addition to these correction coils,
27 ELM control coils with 27 independent power supplies will be used for identification of the error
field structure and, in scenarios when these coils are not used for ELM control or in scenarios when
the coil currents required for the ELM control leave sufficient margin, the ELM control coils could
be used for correction of error fields with n = 1, 2, 3.

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Dynamic (feedback) error field correction (DEFC) would make use of the ELM control coils,
assuming that sufficient ‘headroom’ exists above the current required for ELM control. The
following upgrades of the error field correction system can be envisaged:–

increase of voltage on the ELM control coils, if this is deemed necessary;–

if required, increase the speed of the cooling water in the ELM control coils.

Depending on the need, these upgrades could be considered as early as in the shutdown period
between PFPO-1 and PFPO-2.
4.6.4

Control of Resistive Wall Modes

Resistive wall mode (RWM) control and dynamic error field correction via the ELM control coils
utilize the same control system and associated upgrades. Therefore, any upgrades of DEFC utilizing
the ELM control coils can be applied also to the upgrade of the RWM control capability. The only
potential difference between these two control capabilities is the speed of control (with RWM
requiring more rapid alteration of the perturbed field than DEFC).
If the current requirements for ELM control determined by operating experience in ITER plasma
scenarios allow the application of a significant fraction of the current capability of the ELM control
coils to RWM control, the following upgrades of the RWM control system can be envisaged:-

upgrade the power supplies of the ELM control coils (increase of on-load voltage andreduction of the voltage response time) in combination with measures for reduction of the
low frequency noise (for frequencies less than ~600 Hz) in the diagnostic signal, dBpoloidal/dt,
used for RWM control;-

if required, increase the speed of the cooling water in the ELM control coils.

Depending on the need, these upgrades could be considered prior to the development of DT hybrid
and non-inductive operational scenarios.
Algorithms for the assessment of the fatigue lifetime of the ELM control coils, based on the
waveforms of currents flowing in the coils, should be developed. The residual fatigue lifetime of theELM control coils should be periodically reassessed based on operational experience gained in
ITER scenarios using ELM and/or RWM control.

4.7 Upgrades required for maintaining long-pulse operation
According to the Project Requirements [ITER_D_27ZRW8, 2014], “The extended burn duration
under the hybrid and non-inductive operations may be accomplished with additional investment for
auxiliary systems”. A critical element in providing access to long-pulse operation is the H&CD
capability, and the upgrade options have been summarized in section 4.1. A detailed review of theanalyses of candidate non-inductive plasma scenarios in ITER, including the influence of the
H&CD capability on pulse duration and fusion performance, is given in Appendix E, while options
for upgrading of the H&CD systems are discussed in detail in Appendix G.
An initial assessment of the implications for system performance associated with the requirement tosupport very long pulses has highlighted that a more detailed analysis of system performance
should be performed in advance of Q ≥ 5, 3000 s operation, based on experience from early ITER
operation. Systems that should be analyzed in this respect are: (i) the diagnostic neutral beam; (ii)magnetic measurements and possible back-up measurements should magnetics become unreliable
on such long timescales; (iii) fuelling and pumping systems; and (iv) cooling water system (pulse

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length and pulse repetition rate needs to be assessed for long-pulse scenarios taking into account theseasonal temperature variation and the heat transfer from the basin due to water evaporation andother heat loss by radiation and convection).

4.8 Upgrades for exploration of Demo plasma regimes and for Demo technology
demonstration in ITER
During the FPO phase, ITER can be expected to deploy improved plasma heating and current drive
schemes, as well as components to test first wall and heat extraction technologies, which would besuitable for initial deployment on Demo. These would comprise evolutionary updates to existing orplanned heating and current drive systems to provide greater flexibility, reliability and survivability
and/or higher power density. ICRF launchers, for example, would employ advances in materials
and cooling concepts that are suitable for Demo (e.g. gas cooling) and adapt these approaches toadditional requirements (such as electrical conductivity) to produce rugged designs. Alternative
ICRF launcher approaches such as the folded waveguide design could be tested in ITER as Demo
prototypes. Launchers for high-frequency microwaves used in ECRH and EBW heating and ECCD
have different requirements and opportunities. The survivability of copper mirrors used in thepresent ITER design is a significant concern. An alternative is to implement remote steering
utilizing angled injection into an oversized-corrugated waveguide. The ohmic loss on thiswaveguide is very low so that it can be made of any metal, including a high resistivity metal, suchas tungsten-molybdenum. A remote steering upgrade would serve as a prototype for Demo, or it
may be required if operational experience shows that the present front steering mirror concept is notreliable.
Such H&CD upgrades should bring the level of coupled power to the plasma to the maximum of~110 MW and, together with the upgrades for fuelling and first wall components described above,
would make ITER an ideal test bed for integrating the technology and plasma regimes foreseen in
Demo. Such upgrades would allow ITER to explore plasma operation with Demo-like heating andcurrent drive schemes and first wall divertor materials under Demo-relevant plasma conditions:high Ploss/PL-H, high N, high Prad/Ptot, high <ne>/nGW, high Q (≥10) and steady-state Q (≥5) with
Demo-like H&CD, etc., albeit not simultaneously and/or not at the highest currents. Given theuncertainty in the foreseen H-mode threshold in ITER, an increase of the heating power level may
also be required to access very high Q ≥ 10 regimes in 17 MA inductive operation, which would
allow the study of quasi-self-sustained thermonuclear plasmas in ITER.
ITER can also make significant contributions to the testing of advanced Demo-relevant first wall(not yet included in the first 4 modules of the TBM testing program – see section 3.2) and divertor
materials and structures together with associated heat extraction technologies in an integrated fusion
environment. Such tests would focus on studying the synergetic effects of neutron irradiation andhigh plasma fluxes on the sputtering of advanced materials and in the tritium uptake, retention anddiffusion through high performance helium-gas cooled first wall and divertor panels operating atelevated temperatures (higher than 600 C). Attractive advanced plasma-facing/structural material
combinations, such as tungsten joined to oxide dispersion strengthened low activation steel, could
be tested in a mid-plane port or in a dedicated divertor sector that would expose test samples tovarious degrees of plasma impingement (and heat flux) under controlled conditions.

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Table 4-1(a) – Overview of potential system upgrades at each phase of the IRP

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Table 4-1(b) – Overview of potential system upgrades at each phase of the IRP

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Table 4-1(c) – Overview of potential system upgrades at each phase of the IRP

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Table 4-1(d) – Overview of potential system upgrades at each phase of the IRP

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5 Research Program Accompanying Construction
5.1

H-mode issues

The research in present and near future tokamak experiments (ASDEX-Upgrade, DIII-D, EAST,
JET, JT-60SA, KSTAR, WEST, etc.) should be focussed towards H-mode research in plasmas withthe specific features of the ITER H-mode plasmas, in order to understand and develop validated
models that can be used to predict plasma behaviour in H-mode scenarios in ITER. This involves
not only the study of stationary conditions but also of the access/exit phases to/from H-modes,
where plasma behaviour is already more difficult to control in present experiments, particularly
with W PFCs. In ITER this is further complicated by the relatively low margin of P sep/PL-H and theneed to maintain ELM control in these transient phases.
The main H-mode issues that should be the focus of H-mode research to improve ITER H-mode
performance and to refine the ITER Research Plan are:Characterization of H-mode plasmas (core, pedestal and overall confinement) in ITER-like
conditions, chiefly with electron dominant heating, low torque input, low core particle
source and peripheral pellet fuelling;Characterization of the evolution of the pedestal and core plasma parameters (density,
temperature, impurity concentrations, etc.) in the L-H and H-L transition and developmentof schemes to control them based on ITER-like actuators (particularly heating and current
drive, fuelling, impurity seeding, etc.);

Determination of the plasma parameters that determine the L-H transition in ITER-like
conditions (e.g. low neutral core particle source) and development of schemes to decrease
the required additional heating power to access and sustain high confinement H-modes;

Characterization of the effects of impurities and gas fuelling, required for radiative divertor
conditions in ITER, on the plasma edge of H-modes with metallic PFCs and on divertor
power loads and development of schemes to improve H-mode performance in radiative
divertor conditions that are extrapolable to ITER;Development of ELM controlled/suppressed H-mode scenarios by application of 3-D fields
with ITER-like plasma features (electron dominant heating, low torque input, low core
particle source and peripheral pellet fuelling) including current ramp-up/down phases (i.e.
control of ELMs by 3-D fields with varying pedestal parameters and q95);

Determination of the impacts of ELM control with 3-D fields on H-mode thermal, particle
and fast ion confinement, impurity exhaust and divertor radiative conditions anddevelopment of schemes to optimize them for integrated H-mode operation with ITER-likeactuators;

Characterization of the (core-pedestal) physics processes that determine the changes of Hmode plasma confinement with increasing  for q95 = 3 – 6, and development of
optimization schemes with ITER-like actuators. While this is of importance for the Q = 10
inductive goal, it is essential for the Q = 5 long-pulse and steady-state goals which rely on
H-mode plasmas at q95 > 3.5 and H98 > 1 confinement.

The experimental research in the areas described above must be accompanied by equally intensemodelling and theoretical development programs to provide:282

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physics basis on which to extrapolate the findings in present tokamaks to ITER; and,validated models on which to design ITER plasma scenarios for the ITER Research Plan.

In addition to these overall research issues, specific topics have been identified for each of theresearch plan phases where H-mode experiments are performed; these are described below.
5.1.1

H-mode research related to PFPO-1

Additional research issues of importance for this phase are:Characterization of the effect of TF ripple on the L-H power threshold, pedestal and core
H-mode parameters and ELM characteristics in ITER-like H-modes. Most of theexperimental H-mode plasmas on which ITER H-mode performance projections are based
are NBI heated and, on this basis, significant pedestal degradation (20 - 30%) is expected for
1.8 T plasmas with 1.3% ripple, as discussed in Appendix B. However, it is not yet known if
the same behaviour is expected for RF dominated plasmas with electron heating as the ITER
plasmas will be;Further studies of ICRF plasma coupling at large antenna-separatrix distances and withsecond harmonic H majority and minority (in He plasmas) schemes. The former is important
to determine the way to maintain coupling in ITER hydrogen H-modes at this stage if the
ICRF antenna is available and ripple needs to be reduced to ~0.5%, as this requires a plasma
separatrix-antenna distance of ~ 55 cm in ITER. The latter determines whether ICRF heated
H-modes will only be viable in H or also in He plasmas;Studies of L-H transition and H-mode performance in mixed H+10% He plasmas. This canopen a wider operational window for H plasmas in ITER in PFPO-1 and, more importantly,
in PFPO-2. Potentially, it may allow performing all H-mode programs in PFPO-1 andPFPO-2 with plasmas of this type rather than pure He H-modes.

5.1.2

H-mode research related to PFPO-2

Besides the specific research for PFPO-1 also relevant for PFPO-2, other research issues ofimportance for this phase are:

Validation of shine-through loads in hydrogen and helium plasmas with NBI at high ion
energy ENBI ≥ 500 keV. This can be performed in JT-60SA in advance of ITER PFPO-2
operation and it is essential to determine whether hydrogen plasmas at 2.65 T have a
sizeable operational space in H-mode in ITER;



Validation of modelling for fast particle losses with 3-D fields for ELM control in ITER
with present experiments for a range of 3-D field spectra and NBI ion injection energy.
This is required to determine the way to optimize the spectra applied to obtain ELM
control while minimizing fast particle losses from NBI;Validation of the single pass absorption, power deposition profiles and possible orbit lossesof accelerated minorities for ICRF heating with the three-ion scheme (H-He4-He3)
proposed to heat ITER hydrogen plasmas at either 3.0 or 3.3 T.

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5.1.3

H-mode research related to deuterium and trace T plasmas in FPO

In addition to the general H-mode physics R&D above and the specific issues in PFPO-1 andPFPO-2 which are of relevance for the D phase, the following issues would benefit from dedicatedR&D in present experiments to refine the operational plan:

Evaluation of the need for ECRH to assist plasma breakdown in ITER at 1.8 T. Ideally, thisshould be performed in devices operating with high-Z divertor/PFCs and with a highly
conducting vacuum vessel (plus ITER-like 1.3% TF ripple at the outer midplane) and scaled
current/voltages in the PF systems to mimic the expected conditions in ITER. This isrequired to determine whether the D H-mode development path from a current level of
5 MA to 7.5 MA in this phase should follow a constant q95 approach or a constant
Bt = 2.65 T approach;Assessment of effectiveness of W accumulation control versus location of the power
deposition profiles and electron/ion heating in devices operating with high-Z divertor/PFCs.
This is important to determine the operational strategy to proceed upwards in current andfield from 7.5 MA/2.65T in D and DT H-mode plasmas in FPO.

5.1.4

H-mode research related to hybrid and steady-state D plasmas in FPO

The specific issues to assess by R&D for the refinement of the research plan for hybrid and steadystate D plasmas in FPO are:

Evaluation of the ITER q95 = 4 hybrid and q95 = 5 steady-state target scenarios including the
optimization of current drive to achieve these target scenarios with real-time control
schemes;

Development of validated models (on the experimental assessment above) able to performaccurate simulations for ITER to define the required real time control schemes, as well as topredict the expected performance of these plasmas in ITER. This can provide important
information on the heating and current drive upgrades necessary to achieve the high noninductive current fraction, high-Q plasma scenarios required for the long-pulse and steadystate goals in ITER.

5.1.5

H-mode research related to inductive DT plasmas in FPO

In addition to the general H-mode physics R&D above and the specific issues in PFPO-1, PFPO-2
and D H-modes which are of relevance for the DT phase, the following issues would benefit fromdedicated R&D in present experiments to refine the operational plan:Improvement of the physics basis for T transport in mixed DT H-mode plasmas throughdedicated experiments and modelling. Both isotopic effects on anomalous and neoclassicaltransport need to be understood. In particular the role of core neoclassical transport indetermining the central T concentration for plasmas with low DT central source needs to be
determined;

Improvement of the physics basis for DT effects on W transport and core W accumulation.
This is in part linked with the above issue and to the level that the central levels of DT
density peaking are set by neoclassical transport in mixed DT plasmas. This would make theexperience on W control gained for D plasmas in ITER not to be applicable to DT;Development of robust H-mode access and exit schemes compatible with ITER
requirements that take into account the effect of plasma parameter evolution on edge power284

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flow to simulate the changes to -heating in ITER (e.g. by real-time variation of a fractionof the additional heating power with plasma energy in the H-mode access and exit phases);


5.1.6

Determination of the physics processes that determine edge stability limits and overall
H-mode confinement in conditions with low grad-n in the pedestal and evaluation of theirrelevance to ITER and of the expected H-mode confinement in these conditions.
H-mode research related to hybrid and steady-state DT plasmas in FPO

The specific issues to assess by R&D for the refinement of the research plan for hybrid and steadystate DT plasmas in FPO are:

Development of control schemes required for hybrid and steady-state scenarios (such as ,
current/pressure profile, edge-core integration, etc.) and demonstration of their application
to optimize performance in H-mode plasmas with a large non-inductive current fraction
including: the development of access route to these high confinement/large non-inductive
current fraction H-modes and demonstration of their compatibility with metallic PFCs
(including divertor power load control);

Identification of the conditions required to obtain H-modes with high q95, high H98 (~1.6),
and characterization of the properties of these plasmas in these conditions in a wide range ofexperimental devices. The objective here would be to obtain a relevant experimental
database of these H-mode plasmas to improve the physics basis for the H98 ~ 1.6, Q ~ 5
steady-state scenario in ITER.

5.2 Plasma-wall interaction issues
Section 2.2 has treated at length the most important plasma boundary physics and plasma-wallinteraction issues in the context of the challenge that ITER will face and the areas which will
require particular attention on ITER. R&D required in the coming years was implicitly (often
explicitly) covered in that section and is summarized here briefly in the form of a list of issues
which should be addressed in the years up to the beginning of ITER exploitation. It is worth
mentioning that unlike many of the other key physics/operational areas, plasma boundary issues
cover both physics and materials and, as such, require R&D on tokamaks, linear plasma devices andtransient heat pulse devices (e.g. electron beams, plasma guns), as well as advances in theory andsimulation.
Plasma boundary physics:


The inter-ELM near-SOL heat flux channel width (section 2.2.2 and Appendix C): is theexperimental scaling q,near  1/Ip preserved at high current in ITER? The question is far
from resolved, and there are opposing theoretical positions, but if the values of
q,near ~ 1 mm at 15 MA implied by the experimental scaling and neoclassical drift theory
were to be confirmed on ITER, the operational window for burning plasma operation may
be substantially reduced, particularly given that divertor target component shaping
significantly reduces tolerable parallel heat fluxes. The important question appears to bewhether electron filamentary (blobby) turbulence ‘washes out’ the ion drift scale at highcurrent in ITER. Further theoretical development/simulations and new experimental data
(e.g. additional measurements in JET at high power/high current) are required to resolve thiskey issue.

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Equally, the study of what physics determines the far-SOL power and particle widths andthe possible link to detachment (divertor/SOL collisionality) should continue experimentallyand theoretically. What is the driving mechanism for filament formation, where and at whatfrequency will filaments form on ITER and how will their impact on Be first wall surfaces
determine the net erosion source? It currently remains unclear whether the broad far SOL
profiles seen universally in L-mode at higher density in current experiments will persist in
H-mode at ITER. Recent H-mode experiments on JET, for example, find that broad SOL
density shoulders are not typically seen for vertical target configurations, but are seen forouter strike points on the bulk, quasi-horizontal tungsten target [Wynn, 2018]. This calls into
question the leading explanation for the formation of broad shoulders, i.e. the strong
correlation seen with divertor plasma collisionality [Carralero, 2015]. It may instead belinked rather to plasma-neutral interactions in the divertor target regions.



The recent experimental scaling for the peak outer target ELM parallel energy density, || foruncontrolled ELMs (section 2.2.3 and Appendix C) is a favourable development for ITER in
the sense that H-mode experiments with unmitigated ELMs may be possible at higher
currents than previously thought (subject to satisfactory control of tungsten influxes). To
date, the same scaling dependence has also been found for mitigated ELMs. Moreexperimental work is required here in the coming years to expand the databases for bothcontrolled and uncontrolled ELMs, for larger numbers of devices and for a wider range of
operating scenarios. An improved understanding of the scatter in the scaling database, and inparticular in the dependence ||  (WELM) (with  in the range 0-1) is required. More data
are also required for the inner target on additional devices since, as a result of the ITER
target design and magnetic geometry, this may be the location which sets the limit on
allowable transient energy density. The dependence of || on pressure at the pedestal top
means that mitigating ELMs will still not be sufficient at the highest performance in ITER,
so that either full suppression, or H-mode regimes with naturally smaller ELMs and good
confinement (including the ability of highly dissipative divertors for ELM energy
dissipation, provided the incoming energy pulse is low enough), is required. Exploration ofthese approaches should therefore continue in current devices, as should theoreticaldevelopment and simulation to consolidate the experimental scaling findings.

An outstanding issue is the degree to which Ne seeding will be possible for detachment
control on ITER (section 2.2.5). From the simulation point of view, ITER is large andpowerful enough for N2 and Ne to be roughly equivalent, yet thus far on most of today’s
devices, Ne enrichment is too low for it to be a useful divertor radiator. Here, higher power
experiments on JET with the relevant ITER wall material mix will be key. Developments intheory and simulation must continue along the lines already being successfully pursued to
fully include drifts and a proper description of impurity transport in the non-trace limit into
plasma boundary simulations (see e.g. [Sytova, 2018]). This will also provide more realistic
predictions for the likely target power loading asymmetries in ITER at high power and abetter description of the SOL flows, so important for the determination of migration
patterns. In parallel, the 3-D simulation capability must be developed, such that the physics
model (drifts, recombination, etc.) attains similar maturity to that of the 2-D codes, in orderto properly simulate the likely power loading in the presence of magnetic perturbations for
ELM control, particularly in the dissipative limit, with strong impurity seeding. The very
general issue of proper detachment modelling is a central theme here (together with the
problem of narrow heat flux channels – see above), since the inclusion of divertor target
shaping means that stronger detachment will be required for given incoming parallel heat286

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flux, particularly at the outer target, compared to an unshaped divertor target. All theseelements ultimately combine during the development of reliable and routine detachment
control strategies (section 2.2.9). This is an area which must be assigned higher priority incurrent devices in the approach to ITER operation, paying specific attention to the pursuit of
methods deploying the same sensors and actuators as will be present on ITER. In particular,
the very extensive IR coverage of first wall and divertor surfaces in ITER requires that
operating tokamaks devote more effort to the use of IR for heat flux control.


The castellated ITER divertor target design has stimulated a great deal of recent R&D
focussed on the power loading of component edges (section 2.2.7). Calculations, validated
in part on present machines, show that this gap edge loading (Appendix C) may set a lower
limit on allowable energy loss per ELM, and it seems difficult to avoid this by shaping
design. Experiments should thus continue to study edge power loading and the effect offinite ion orbits to build more confidence that the extrapolations to ITER may be trusted.

Plasma-materials interaction:


Loss of material strength during tungsten recrystallization has been shown to favour theappearance of macro-cracks which can propagate down to the cooling tube. To avoid thesedeleterious effects, an operational budget will need to be defined which takes into accountthe fact that recrystallization is a time- and temperature-dependent effect. To date, therecrystallization kinetics of the tungsten material to be used in ITER is not wellcharacterized and detailed studies are needed to allow a reliable operational budget to be
defined for ITER operations.



Even if deep recrystallization is avoided by proper control of the steady-state divertor heat
fluxes, surface roughening/cracking/melting can occur, depending on the ELM energy
density. The evolution of the surface damage and its dependence on the base material
temperature have been studied in detail, and it is clear that the damage threshold is very low
for ITER. One big unknown for ITER, however, is whether and how such surface damage
might affect the material compatibility with high performance plasmas. In particular, it is atpresent impossible to define a tolerable level of damage before the degraded material
properties would start to have an impact on plasma performance because of enhanced
tungsten release or reduced power handling capabilities. This is clearly an area where further
research is needed before the start of ITER operations.



The effect of long-term stationary and transient plasma exposure (be it H, D, DT or He) on
the thermo-mechanical properties of W clearly remains an important topic where little
information is available. Although indications exist of material property modifications
following plasma exposure, extrapolations to the very high fluence expected in ITER are
currently hazardous. Results from high fluence (>1028 m-2) experiments are currently very
scarce, especially for the case of mixed species (including extrinsic seeded impurities),
although the recent development of high flux steady-state plasma devices should expand theexisting database. These experiments will need to be accompanied by detailed mechanicalmeasurements allowing the quantification of changes in material properties such as ultimate
strength, ductility, etc., which will require a systematic use of micro-mechanicalmeasurements. Understanding long-term plasma exposure effects will also require thedevelopment of multi-scale material modelling capable of describing the evolution of thematerial properties at different temporal and spatial scales, therefore bridging the gap fromthe plasma-surface interaction processes (occurring close to the surface) to the long range
diffusion of D/T/He atoms and their effect on the material bulk.
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Related to the above, the existence and importance of synergetic effects should be
investigated in more detail in the coming years. This includes, for example, the possible roleof plasma-induced defects (voids, bubbles) on the recrystallization kinetics of the tungsten
material. Recent investigations have indeed revealed a possible retarding effect of plasma
exposure on recrystallization. An additional issue is the effect of stresses created by the highlocal concentration of plasma species in the near-surface region on the material thermal
shock behaviour.



Systematic analysis of the JET bulk tungsten divertor lamella would provide important
information for ITER. Even if the conditions in JET (inertial cooling, different stress
distribution) are quite different from those in ITER, the tungsten material in JET has been
exposed to significant plasma fluence in the presence of a large number of ITER-relevant
ELMs, providing unique information about the effect of long-term exposure to the tokamak
environment on tungsten.

Formation of tritiated ammonia might represent an issue for the duty cycle of ITER
operations, since high temperature regeneration of the cryopumps would be needed torecover ammonia (and the tritium within) from the active charcoal used in ITER for H/D/T
and He pumping. Ongoing studies of ammonia formation rates during nitrogen injection
should be continued and complemented by edge plasma simulations where thenitrogen/hydrogen chemistry is accounted for. The recent development of a reaction
database for the SOLPS-ITER code is a first step in this direction, but would benefit fromadditional data on surface reaction rates, which are crucial since ammonia formation mainly
proceeds through surface reactions. In parallel, the possibility of preventing ammonia
formation during nitrogen seeding should be explored in a fashion similar to the so-calledscavenger effect, which was promoted to mitigate the formation of tritium-rich carbon codeposits.

5.3 MHD stability issues
A number of MHD stability control issues continue to be the subject of active research during the
ITER construction phase. The Plasma Control State-of-the-art and R&D Plan [Snipes, 2016-2] fromthe PBS-47 Plasma Control System Preliminary Design Review gives a good overview of recent
R&D and suggestions for future R&D for all of the plasma control functions, including MHD
stability control. A brief summary of the recommended MHD stability control R&D is given below.
5.3.1

Disruption characterization, prediction and mitigation

Specifying mitigation success rates for ITER disruption management requires more input on thecharacterization of electromagnetic loads as well as on the energy deposition distribution during
runaway electron impact. The ITPA disruption database shows that some tokamaks encounterfrequently halo current amplitudes at levels of ITER load category III [Eidietis, 2015]. However, it
is also seen that machine specific ‘hidden’ parameters appear to determine the halo current fraction
and the toroidal peaking. This has to be further understood to improve predictions for ITER. During
the ‘progressive start-up’ approach, EM loads will be monitored and models will be validated as
plasma parameters are increased in the course of the implementation of the Research Plan. A better
understanding of the physics of asymmetric VDEs is required to ensure that these can be assessed
appropriately in applying this approach. Models describing these events have to be assessed andvalidated through experiments. Initial modelling has been performed to address the impact of
runaway energy deposition on the first wall or divertor components to specify critical plasma288

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currents for which runaway formation would represent a significant risk for water leaks [Lehnen,2016]. A full assessment requires more information on RE energy, the deposition process and theresulting energy load distribution.
The disruption predictor inside the PCS aims to trigger the DMS within its reaction times [Pautasso,
2016]. High levels of prediction success will have to be achieved for high energy plasmas [Lehnen,2016]. Predictors currently under development are either physics-based threshold approaches or
statistical methods using training data sets. R&D in this area is required to extend threshold
approaches, such as that described in [de Vries, 2015], to reach the required success rates.
Disruptions in which the prediction algorithm has failed in current devices have to be assessed withrespect to their severity. In this respect, it is important to identify the expected disruption types in
ITER and to assess those critical for disruption mitigation. Training and testing data sets have to be
selected such that they reflect the expected types of disruptions in ITER. The threshold-based
predictor will be used from the beginning of operation and may be complemented at a later stage ofthe project with a data-driven predictor. All of these approaches have to ensure portability andscalability between devices to avoid excessive commissioning/training in ITER and also to reduce
the risk of insufficient prediction success rates when progressing in the operational space and whentransiting to new operational scenarios (e.g. from L-mode to H-mode).
The ITER strategy for establishing reliable and effective disruption mitigation is based on a stepwise approach to account for the gaps in the present physics and engineering basis. Most urgent
R&D has to focus on physics and technology issues related to the baseline DMS that will use
shattered pellet injection (SPI) systems distributed across several ITER port plugs to effect thermal,
electromagnetic and runaway electron load mitigation at disruptions. Given the current
understanding of the challenges in implementing a DMS with the required level of reliability in
ITER, extensive R&D is required to provide the basis for a timely upgrade of the baseline systemshould results from the ongoing R&D, or, subsequently, from ITER operation, reveal that thebaseline specification does not adequately meet the mitigation requirements. A final element of the
R&D program for ensuring that ITER will have a satisfactory disruption mitigation capability at thehighest plasma performance is the exploration of alternative approaches to disruption mitigation.
These could be related to more efficient material delivery to the plasma or to modifications of thepresent mitigation scheme, which is primarily based on energy dissipation through radiation. This
provides an element of risk mitigation for the baseline implementation of the DMS concept (i.e.
SPI) given the critical role played by the DMS in the successful implementation of the IRP.
The two main outstanding issues in understanding the physics of disruption mitigation are thesuppression or avoidance of runaway formation during the disruption mitigation process (i.e. during
the initial intervention aimed at mitigation of thermal and electromagnetic loads) and – as a second
layer of defence – the dissipation of runaway energy should a runaway beam have formed in spiteof the initial intervention. Therefore, physics R&D needs to focus on improving the understandingof all processes that drive RE seed formation, including the role of the thermal quench MHD and
the impact of the impurity source distribution (determined by, e.g., pellet fragment ablation).
Similarly, physics R&D needs to address all mechanisms that impact on the efficiency of the energy
dissipation scheme, such as the runaway interaction with impurities, the impurity penetration
process and the evolution of the runaway beam, including possible instabilities.
The R&D needs around the implementation of the DMS baseline concept can be divided into threegroups: a) integration of the system in the port plugs, b) the pellet generation and injection process,and c) the effectiveness and robustness of the system during ITER operation. Addressing theseareas will require experimental work in laboratories as well as at tokamaks together with modelling289

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and theory development. The work related to groups a) and b) is very much technology oriented andincludes, for example, adapting the SPI technology to the specific restrictions in ITER on space andport plug availability, but also to addressing questions specific to SPI, such as optimizing pellet
release and the shattering process to ensure high reliability and effectiveness of the system.
Work related to group c) focusses on matching the SPI parameters to the needs for most efficient
mitigation. This includes addressing the optimum size distribution of the pellet fragments, theoptimum pellet velocities and injection directions, as well as the effectiveness of multiple pellet
injection. Finally, extensive tests at present devices with mitigation systems representative of the
final design choice for the ITER baseline system will be required to minimize the risk that the
mitigation performance is insufficient at some point during ITER operation when approaching highcurrent or high energy operation. Currently, SPI experiments are performed at DIII-D and are inpreparation for JET. These experiments focus on narrowing down requirements for the DMS, suchas injection angles, effectiveness of multiple injections, etc. Priority should be given specifically to
R&D on the validation of the concept of material injection to avoid runaway formation and on
confirming the required injection quantities, sequences, and species. The effectiveness indissipating the runaway energy as a second layer of defence has to be further addressed as someexperimental results [Reux, 2015; Commaux, 2016] and recent modelling work [Konovalov, 2016]
have raised questions on the viability of this strategy. Radiation heat loads can be an issue for
mitigated disruptions with high thermal energy. To date, radiation asymmetries have been assessed
only for MGI mitigated disruptions [Izzo, 2013; Shiraki, 2016; Lehnen, 2015-2] and must be
assessed for SPI. Theory and modelling work is needed to conclude on the required injection
quantities and shard sizes, and on the allowable fraction of gas and liquid during SPI. Here, it isimportant to account for the various plasma scenarios for which mitigation is required - this canhave a significant impact on ablation rates and thus assimilation efficiency.
5.3.2

Sawtooth control

It is important to extend the methods that have been demonstrated to be effective for controllingsawteeth in present experiments by further exploiting localized current drive or fast ion generation
supported by model validation that would allow application in ITER. Requirements for heating andcurrent drive actuators for sawtooth control have been described in detail in an ITPA report[Chapman, 2011]. While ECCD has been shown to control sawteeth effectively for decades, only
recently have such demonstrations been replicated in the presence of core energetic particles.
Sawtooth destabilization of long period sawteeth induced by ICRF generated core fast ions withenergies ≥0.5MeV was achieved in Tore Supra, even with modest levels of ECCD power[Lennholm, 2009a; Lennholm, 2009b]. More recently, sawtooth control using ECCD has been
demonstrated in ITER-like plasmas with a large fast ion fraction, wide q=1 radius and longuncontrolled sawtooth periods in DIII-D [Chapman, 2012]. As expected from simulations, thesawtooth period is minimized when the ECCD resonance is inside the q=1 surface. Sawtooth
control using driven current inside q=1 allows the avoidance of sawtooth-triggered NTMs, even atmuch higher pressure than required in the ITER baseline scenario [Chapman, 2010; Chapman,2013-2]. Operation at βN=3 has been achieved in ITER demonstration plasmas – which otherwise
often experience sawtooth-triggered 2/1 or 3/2 NTMs – when sawtooth control is applied using only
modest ECCD power [Chapman, 2012]. Such avoidance of NTMs permitting operation at higher
pressure than otherwise achievable by application of steerable core ECCD sawtooth control has also
recently been demonstrated in ASDEX Upgrade [Chapman, 2013-1].
TCV has demonstrated feedback control of the sawtooth period by actuating on the EC launcher
injection angle in order to obtain the sawtooth period at a predetermined value [Paley, 2009]. Fine290

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control over the sawtooth period has been demonstrated on Tokamak à configuration variable(TCV) using either `sawtooth pacing' via modulated ECCD with real-time crash detection[Goodman, 2011], or `sawtooth locking', where the sawtooth period is controlled even in theabsence of crash detection in a reduced region of duty-cycle vs. pulse-period parameter space[Lauret, 2012; Witvoet, 2011]. The applicability of these control methods to ITER has been tested[Kim, 2014] and sawtooth locking experiments have been performed on KSTAR [Kim, 2015;
Jeong, 2015] and AUG [Kim, 2015] in the presence of energetic particles.
Control of sawteeth by ICRF in the presence of core energetic particles has been widely exploited
on JET, e.g. [Eriksson, 2004; Eriksson, 2006; Sauter, 2002]. The fundamental difference between
ECCD and ICRF for sawtooth control is that ECCD modifies the local magnetic shear at the q=1
surface, while the ICRF kinetically destabilizes the kink mode. Furthermore, ICRF control has also
been demonstrated in plasmas with significant heating power on-axis from neutral beam injection
and high βp well above the critical threshold for triggering of 3/2 NTMs in the absence of sawtooth
control [Coda, 2007]. Subsequently, it was noted that the sensitivity of sawtooth destabilization by
ICRF was very high and effective sawtooth control required an accuracy of the resonance position
with respect to the q=1 surface of less than 0.5% (i.e. within 1 cm of the q=1 surface in JET), farmore sensitive than expected from a control mechanism involving a modification of the magnetic
shear. Recent JET experiments using 3He minority heating (so that the driven current was
negligible) on the high field side just outside the q=1 surface lead to a strong destabilization for
counter-propagating waves (-90°) and a strong stabilization for co-propagating waves (+90°)[Graves, 2010]. This sawtooth control scheme via affecting the kink mode potential energy has
subsequently been demonstrated in H-mode plasmas with significant core heating too [Graves,2012], adding credibility towards its applicability in ITER. Finally, real-time control throughvariation of the ICRF frequency has been attempted with some success on JET [Lennholm, 2011],
though the frequency variation is much slower than anticipated in ITER.
A detailed experimental program on sawtooth control in ITER should be developed arounddedicated R&D on existing devices, performed with the goal of optimization of the commissioningof sawtooth control for ITER. Experimental activities should include minimization of the number ofthe dedicated shots, assessment of required duration of the plasma current flattop, q95, etc.,
optimization of the control methods and algorithms, and development of requirements fordiagnostics at each phase of the IRP. The corresponding R&D should be carried out using
diagnostics and actuators similar (to the extent possible) to those in ITER.
5.3.3

ELM control

For plasma currents exceeding 6 – 9 MA, depending on the assumed wetted area of the heat flux
[Loarte, 2010], type-I ELMs must be controlled or mitigated in ITER to avoid excessiveintermittent heat load and subsequent material erosion from the first wall blanket modules anddivertor target plates; at high plasma performance, ELM control is critical for machine protection.
This edge instability is believed to be an interplay of low to intermediate mode number peelingballooning modes and high mode number kinetic ballooning modes whose onset is well predicted inexisting devices by the EPED model [Snyder, 2011].
It has been found experimentally that the ELM amplitude decreases with increasing ELM frequency[Lang, 2003]. ELM pacing has been demonstrated on several machines with deuterium pellets[Lang, 2004; Lang, 2011; Lang, 2013] and with lithium granules [Mansfield, 2012]. Deuterium
pellet injection at up to 64 Hz can potentially be used in ITER to induce ELMs frequently enoughso that the ELM amplitude remains relatively small. This frequency is expected to be sufficient to291

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reduce ELM perturbations to acceptable levels with WELM < 0.7 MJ for the Q=10, 15 MA
inductive scenario [Loarte, 2010]. However, the pellet size must be chosen to have sufficient
penetration to trigger the ELM and yet have little impact on fuelling. Resonant magnetic
perturbations have also been shown to stabilize or mitigate ELMs under some conditions in existing
devices [Evans, 2008; Suttrop, 2011]. The in-vessel ELM control coils will be used to actively drive
an n = 3 or n = 4 edge magnetic perturbation to stabilize or mitigate ELMs. An ELM control coilcurrent of up to 90 kAt will provide a magnetic perturbation in the edge pedestal region of up to
6.6×10-4 × 5.3 T, which is expected to be sufficient to stabilize ELMs based on extrapolation fromexisting devices [Loarte, 2010]. To better distribute the heat load, the ELM control coil perturbation
will be capable of rotating at up to few Hz at full coil current.
5.3.4

Neoclassical tearing mode control

Neoclassical tearing modes represent a significant challenge for present tokamaks and are expected
to be even more important in ITER. Sawtooth and ELM control will help to avoid creating seedislands for NTMs [Sauter, 2002]. The most effective means of NTM control has been demonstrated
with localized ECCD inside the magnetic island, as demonstrated in DIII-D [La Haye, 2002;
Humphreys, 2006] and ASDEX Upgrade [Günter, 2003; Zohm, 2007]. The precise specifications
for an effective NTM stabilization system with ECCD on ITER are still an active area of researchwithin the fusion community [La Haye, 2009; Sauter, 2010]. The response time to initially turn on
EC to full power is 10 ms and to turn EC power on from standby state is 1 – 10 ms. The timerequired to begin sweeping the UL mirrors following a request by PCS is ~200 ms and the sweep
can cover one third of the plasma cross-section in ~3 s (~6º/s). This sweep rate should be adequate
for NTM control to move between the q = 1.5 and q = 2 surfaces [La Haye, 2009]. The EC fullpower modulation will be available with a frequency up to 5 kHz, with a variable duty cycle.
Modulated ECCD with absorption and current drive in the O-point of the islands only (ECCD
power off for X-points) can be yet more effective than CW ECCD, if island widths are substantiallynarrower than that of the ECCD as demonstrated in ASDEX-Upgrade [Maraschek, 2007]. However,the change in the ITER ECRH launcher mirrors to front steering, which was implemented during
the design review, allows narrow ECCD deposition, which may obviate the need for modulation sothat CW ECCD may be effective enough, if well-aligned on the rational surface, for thestabilization of NTMs in ITER.
Techniques to allow a ‘saturated’ island to occur and then turn on and align ECCD for stabilization
are inadequate in ITER where 2/1 islands must be kept small to avoid locking. However, someexperiments have been successful in avoiding or quickly stabilizing NTMs when they first start togrow [Nagasaki, 2003; La Haye, 2005]; this is the approach modelled for ITER [La Haye, 2009].
After stabilization of an NTM and/or before it ever occurs, it was shown that ‘active tracking’ of thereal-time MSE EFIT location of the q=2 surface and of the location of the ECCD by TORAY,TORBEAM or a refraction analogue based on interferometers, in order to keep alignment goodenough between the ECCD driven current and the q=2 location, is a feasible strategy to prevent theNTM modes from reappearing and/or to never occur [Prater, 2007]. However, this strategy for
NTM mode control requires gyrotron power to be applied throughout the high-β phase and thealignment relies on the real-time estimation of the mode location which can have uncertainties.
Recent TCV experiments have introduced a new way to implement this strategy by sweeping
ECCD around the estimated mode location to compensate these possible uncertainties from realtime estimations [Kim, 2015].
Dedicated R&D on existing devices should also be undertaken with the goal of optimization of thecommissioning strategy for NTM control in ITER. This program should include minimization of292

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number of the dedicated shots, assessment of the required duration of the plasma current flattop,
q95, N, etc., optimization of control methods and algorithms, and development of requirements fordiagnostics at each phase of the IRP. The corresponding R&D should preferably be performed
using diagnostics and actuators as similar as possible to those available in ITER.
5.3.5

Error field control

Error field control in ITER is necessary to correct for small misalignments in the CS, PF, and TF
coils and asymmetries in the coil feeders, and to minimize the influence of ferromagnetic inserts,
TBMs, NB Magnetic Field Reduction system, holes and port plugs in the vacuum vessel and othernon-axisymmetric features of the magnetic field topology in ITER, in order to avoid locked modes
and to enable the generation of a local field null for plasma initiation. Such error fields can induce a
torque slowing down the plasma toroidal rotation, which can then lead to a higher incidence oflocked modes and disruptions. Ramps in the n=1 error field produced by external coils can be used
to determine the low density locked mode threshold [Buttery, 1999; Luxon, 2003; Wolfe, 2005;
Buttery, 2012]. Error fields also enhance resistive wall modes (RWMs) at high β and thus theircontrol will help avoid RWMs. The control of error fields and their effects on locked modes havebeen studied for many years on a number of devices [Scoville, 2003; Buttery, 2000; Morris, 1992;
La Haye, 1992; Park, 2007]. Recent dedicated experiments on DIII-D have studied the effects oferror fields due to TBMs particularly on H-mode confinement and plasma rotation [Snipes, 2010].
Three sets of 6 top, bottom, and side external correction coils (CC) will be used in ITER to
minimize error fields within the 320 kAt top and bottom and 200 kAt side correction coil current
limits. In the addition to the correction coils, ELM control coils will be used for identification of theerror field structure and, in scenarios when these coils are not used for ELM control, or in scenarios
when the coil currents required for ELM control leave sufficient margin, the ELM control coils
could be used for error field correction.
Historically, analysis of n = 1 mode error fields leading to mode locking (LM) in ITER has been
performed using two LM criteria. The first is the ‘3-mode’ criterion [Hender, 2007], which wasused for the ITER design since the Engineering Design Activities (EDA) [ITER Technical Basis]
and it was also used in the Project Requirements up to version 5.3 [ITER_D_27ZRW8, 2014]. The
second is the ‘overlap’ error field criterion, which is the criterion on the resonant error field for then = 1 toroidal mode. This was developed around the analysis performed with the IPEC code forthree fiducial ITER plasmas [Park, 2008]. The review of progress made in the last decade in thearea of error fields, summarized in the report of MHD Topical Group of the International Tokamak
Physics Activity (ITPA) [Park, 2017], concludes that: i) the 3-mode criterion must be retired and ii)
as far as the LM avoidance is concerned, only the overlap (resonant) criterion should be used. It
should be noted, that the criteria for error field correction are not final and they are studied in the
ITPA framework. In particular the database of the overlap n = 1 criterion will be increased, andcriteria for correction of n = 1 non-resonant error fields (e.g. effecting plasma rotation and L to H
mode transition), as well as criteria for n = 2 modes remain to be developed.
Optimization of error field control commissioning should be investigated within the R&D programsof existing devices in order to develop a suitable methodology allowing minimization of the time
allocated to this activity in ITER. The program should include minimization of the number of thededicated shots, assessment of required duration of the plasma current flattop, q95, N, etc.,
optimization of control methods and algorithms, and development of requirements for diagnostics ateach phase of the IRP. The corresponding R&D should be undertaken using diagnostics andactuators similar to those available in ITER.

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5.3.6

Resistive Wall Mode control

To achieve high values of Q in steady-state regimes, high values of the normalized pressure, βN, arerequired. Plasmas with low li and high βN may be unstable to Resistive Wall Modes (RWM).
Unstable RWMs can be generated from a plasma initially below the ideal low-n kink/ballooning nowall limit, such as the ITER baseline scenario, by local increases of pressure gradients via transportbarrier formation [Maingi, 2010], or other events that significantly increase the plasma pressure
profile peakedness (e.g. H-L mode back-transition) and thus substantially lower the ideal no-walllimit [Sabbagh, 2004].
RWM stabilization is best accomplished by a combination of passive kinetic stabilization [Betti,
2004] and an active feedback control system capable of sensing low-n RWM activity [Okabayashi,
2001; Sabbagh, 2006]. Calculations using kinetic RWM theory for ITER by two independentgroups (using the MARS-K and MISK codes) show that passive kinetic stabilization is possible in
ITER through the combination of thermal and energetic particle effects (including -particles), butthat the operational space of interest is largely unstable, and thus that passive stabilization must be
supplemented by active control [Liu, 2009; Berkery, 2010]. The studies conducted for ITER using
the MISK code utilize a theoretical model that has been quantitatively benchmarked against the
NSTX and DIII-D experiments.
In ITER, the ELM control coils may be used for feedback stabilization of RWMs with an upgrade
to their power supplies, if necessary. These coils provide a flexible set of actuators capable ofhandling feedback control of multiple toroidal mode numbers up to n = 3, and different poloidal
field perturbation spectra at a given n (multi-mode control at constant toroidal mode number)[Sabbagh, 2010; Sabbagh, 2011]. The power supplies driving the ITER ELM control coils will needto create an AC toroidally rotating field to track RWM rotation during feedback control.
A study of RWM active stabilization in ITER by the ELM control coils was performed with the
CarMa code for two 9 MA plasmas with weak reverse shear having N =2.94 ( ≈ 15 s-1) andN =3.14 ( ≈ 40 s-1). The best achievable performance of the ELM control coils has been computedin terms of the maximum allowable initial perturbations of the vertical (poloidal) magnetic field,
measured by the outboard in-vessel sensors, which is recoverable by any feedback controller withcurrents and voltages inside the design limits of the power supplies. If all three sets of ELM control
coils (Upper, Equatorial and Lower) are used for RWM stabilization, the maximum recoverable
magnetic field perturbation is about 40 mT for both reference 9 MA plasmas. The maximum
recoverable initial perturbation reduces to about 20 mT, if only the Equatorial ELM control coils
are used for the RWM stabilization.

5.4 Scenario development issues
Operational scenarios need to cover all phases of the discharge consistently – the breakdown, rampup, flat-top, burn and ramp-down – with the time evolution providing proper linkage between andduring these phases. Most of the attention in describing ITER operation has focussed so far on theburn phase. But, as known from the experience of operation in present devices, it is necessary to
identify controlled sequences that arrive at the optimum conditions to achieve high performancestationary plasmas even when these plasmas are not dominated by the physics of burning plasmas.
In addition to the constraints imposed by, for example, MHD stability control or plasma-surface
interactions, operational scenarios must link the plasma breakdown, ramp-up, flat-top and rampdown stages in a manner that is consistent with the control systems. Additional constraints may
apply when operating at lower field and current or with a species other than DT fuel (e.g. low
current operation in hydrogen or helium H-mode operation) in ITER that do not make these294

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scenarios any easier to operate due to these additional constraints (marginal H-mode operation,
shine-through limits, etc.), as studied in [Kim, 2016] and also discussed in detail in section 2.5.3.
Research and development activities in the area of plasma control will be a central aspect of theresearch program in the scenario development area. Burning plasma scenarios in ITER will makeextensive demands on the ITER control capability. Limitations on actuator capabilities or capacities
require an integrated control approach, with the possibility of actuator sharing. Demonstrating not
only that the many plasma control techniques that are being developed in the present experimental
programs (e.g. MHD stability control, radiative power control, Internal Transport Barrier (ITB)
control, current profile control) can be turned into reliable and robust control technologies that canbe applied to ITER in an integrated plasma scenario, is a major research activity which willaccompany ITER construction.
It is, furthermore, very important that these experimental studies are accompanied by a substantial
effort in the simulation of ITER operational scenarios. Present day devices will never be capable tofully emulate ITER operational scenarios, but they can validate individual models that can be used
for the predictive simulation of such scenarios. ITER operation is expensive and any component
damage resulting from operational errors can potentially lead to high repair costs and significant
downtime of the facility. Hence predictive simulations, using the ITER integrated modelling
platform (IMAS), to assess plasma control, develop operation scenarios and prepare experiments,
will be essential for reliable and efficient ITER operation [Parail, 2009; Imbeaux, 2015; Pinches,
2016]. IMAS must meet the challenge of simulating complete discharges including all phases fromplasma initiation to termination, prior the experiment, to anticipate their behaviour andperformance, and to tune controls and scenarios accordingly. The ITER integrated modelling
platform shall equally be capable of reconstructing experimental discharges based on the actual
measurements in order to ensure the understanding of all physics aspects related to the experiment.
5.4.1

Breakdown and burn-throughThe large size of ITER brings an immediate challenge, namely achieving effective and reliable
plasma initiation. Although calculations based on the existing tokamak experience suggest thatbreakdown using only the inductive electric field should be achievable, the margin appears to be
low and breakdown assisted by several MW of ECRH, is foreseen. To obtain a greater confidence
that the breakdown has an adequate margin to establish reliable initial plasmas for the ramp-up
phase, detailed modelling of the breakdown using the ITER geometry, PF system, ECRH sources aswell as passive components, such as for example the vacuum vessel, is needed. This modellingshould also include an accurate treatment of the impurity behaviour. Variations, in the first wall(e.g. a steel limiter for First Plasma, and a Beryllium main chamber wall from PFPO-1 onwards),
applied poloidal flux (e.g. the CS precharge) and, most importantly, breakdown at third and half ofthe full ITER toroidal field of 5.3 T, should be also included in these studies.
Progress has been made in this area over the recent years, e.g. by: rehearsing ITER breakdown inpresent devices [Sips, 2009; Jackson, 2010], assessing the effect of a beryllium wall [Kim, 2012-1;
Kim, 2013; de Vries, 2013] as well as use ECRH to assist plasma formation [Jackson, 2011],
culminating in fully simulated ITER plasma initiations [Kavin, 2017]. However, better models of
ECRH assisted pre-ionization and breakdown need to be developed, as well as a better
understanding of the impact of the ITER vessel size with respect to the breakdown area, in order topredict accurately and to optimize plasma initiation in ITER.

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5.4.2

Current ramp-up

The current ramp-up phase raises several operational issues that need to be addressed selfconsistently. The constraints imposed by the PF system during the current ramp-up are
considerable: forming an X-point and controlling its position through the remainder of the current
rise, maintaining the internal inductance, li, in an acceptable range, minimizing resistive losses toincrease the duration of the burn phase, avoiding current, field force, and voltage limits on the coils,
etc. [Sips, 2009; Jackson, 2010]. X-point formation may also have to be done early in the current
ramp-up in ITER using a reliable shape control to avoid the excessive power loads on the beryllium
first wall panels due to narrow power e-folding lengths in the near-separatrix limiter SOL.
A large theoretical and experimental effort has been devoted to improving the understanding ofplasma confinement in present tokamak plasmas and to evaluate the implications for ITER.
However, most of this effort has addressed flat-top conditions in H-mode plasmas while ohmic and
L-mode plasmas always precede the H-mode in the current ramp-up in ITER and dictate the initialconditions for the flat-top, or burn conditions. One of the most critical issues is the evolution of thecurrent profile during current ramp-up, where the radial profile of the thermal conductivity and theresulting resistivity profile determine whether the current profile will be peaked and lead tosawtooth activity (higher li) or broad (lower li) and will more easily connect with high performance
conditions for the advanced scenarios, such as hybrid and steady-state operation. Predictivemodelling of the current profile behaviour during the ramp-up is still not satisfactory[Voitsekhovitch, 2010]. Moreover, the influence of high-Z impurities on this phase should be better
understood [Hogeweij, 2015]. The transition to H-mode conditions, if performed in the current
ramp-up, introduces another dimension to the evolution of the internal inductance through theappearance of the bootstrap current in the pedestal when the plasma enters the H-mode. A low li
from a large edge current increases the demands on the PF system (i.e. PF6) for control of the Xpoint.
A second aspect is the increase of plasma density simultaneously with the current during the rampup phase. This will not be as straightforward in ITER as in present devices, due to the limited
efficiency of edge neutral fuelling [Kukushkin, 2013-2]. This, and the slow response of the ITER
gas fuelling injection system with the valves located far from the vessel, may complicate density
control. Pellet injection may be necessary to assist in fuelling and controlling plasma density in the
ramp-up phase. However, this will be limited for helium operation scenarios, scheduled during
PFPO-1 and PFPO-2, due to the dilution of the helium plasma by the hydrogen pellets. Detailed
modelling of the fuelling of ITER ramp-up scenarios, for all species, as well as assessing theachievable density control, will provide a better insight in the expected density behaviour during
this critical scenario phase [Garzotti, 2012; Ravensbergen 2015; Militello-Asp, 2016].
5.4.3

Current flat-top

As the main phase of a tokamak discharge, the current flat-top usually attracts most attention in
scenario development. In the scope of the analysis of R&D for ITER there are a number of issues
that should be considered: the achievable flat-top duration, the reliable entry to high performance
operation (i.e. access to H-mode operation and eventually entry to burn) and the integration ofvarious control schemes.
A proper assessment of the maximum current flat-top duration, will depend on a number of factors,
the magnetic flux consumed during plasma initiation and ramp-up, the consumption during the flattop, which depend on the plasma resistivity (i.e. related to the plasma temperature and plasma
heating) and on the non-inductively driven current and also on other factors such as the cooling
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capacity of the superconducting coils, and the achievable control of the plasma current density
profile. All these factors will eventually determine the burn duration of the ITER baseline Q = 10
DT as well as of advanced scenarios, but also the available flat-top length for lower current, lower
temperature (i.e. higher resistivity) non-active scenarios to be explored during PFPO-1 and PFPO-2.
The entry to H-mode will need to be controlled with the accompanying change in , which will
require integrated control of the auxiliary heating sources and plasma radial position. The transition
to high performance, especially concerning the transition to burn, requires the gradual application ofcontrol of the steady-state heat flux to the divertor and also the transient loads due to ELMs.
Although heat flux and ELM control techniques have been applied in present devices in stationary
conditions, controlled transitions to high power H-modes are rarely explored.
Controlling the plasma performance, and more specifically kinetic profiles, in present tokamaks has
advanced significantly, which helps to support the case for operation of ITER close to its theoreticaland empirical performance limits [Kim, 2012-2; Felici, 2012]. Under these conditions, plasma
control becomes more sensitive to the desired flat-top plasma regime. During the flat-top the
discharge performance, fuelling, stability and exhaust need to be controlled. Present devices havedeveloped numerous schemes to control sawteeth, NTMs or ELMs and the applicability to ITER
has been assessed. However, there is still a need to develop a reliable methodology to integratethese control schemes into one operation scenario, with high performance together with MHD
stability, kinetic plasma and current profile, fuelling and exhaust control. Larger emphasis should begiven to such integration and verification, identifying how these control schemes may interact witheach other or to identify how they are linked. The latter may impose additional requirements to theindividual control schemes; for example the use of auxiliary heating systems to control MHD
instabilities will impact on the control of the exhaust. The sharing of auxiliary heating sources toeither control MHD instabilities or the kinetic profiles need to be assessed, while the MHD stability
itself may in turn depend on the kinetic and current density profiles. Burn control will also require,besides density control, the control of the fuel ratio. This will require better models for to thefuelling of high performance plasmas, in particular using pellets [Garzotti, 2012].
5.4.4

Plasma termination

The controlled termination of high performance plasmas is an important, but often overlooked,
phase of the tokamak discharge. For ITER, considerable effort will have to be devoted to thedevelopment of appropriate plasma terminations, with special attention to the transition from alargely self-driven burning state to one that is controlled by the auxiliary systems [Sips, 2009]. In
some sense, this mirrors the entry into burn phase, but there are significant differences in the
dynamics, such as the density decay and current profile development, that are opposite. These
differences complicate the control with respect to the entry phase that makes it difficult to avoid
operational limits in many circumstances; this, in the worst cases, may lead to a disruption. This
control problem is exacerbated by the fact that at the end of the discharge, the device is usuallyoperating close to its technical limits (e.g. CS current limit). For unplanned terminations, triggeredby off-normal events, the situation is further complicated. The ability to carry out a well-controlled
termination contributes significantly to the avoidance of disruptions.
The termination phase should achieve a simultaneous ramp-down of the plasma current, kinetic
energy and particle density while maintaining control over the radiation levels, plasma position andshape (i.e. to avoid overheating the first wall) and VS, staying within the capabilities of the poloidal
field coils and power supplies and heating systems. Stability boundaries and general operational
limits must also be avoided. ITER will operate at high densities and a controlled density decay is297

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important to avoid the density limit in the termination phase, while also managing the H-L transition
timing and exit from fusion burn. ITER power supply limitations and the thick low resistivity
vacuum vessel slow the control response for vertical stability (VS) and the radial position that canbe achieved. Increasing the current ramp-down rate leads to higher li, which affects the plasma
vertical stability and its control, while the associated  drop for fast H-to-L back transitions may
compromise the radial position control. A further concern is the control of impurities and radiation
during this phase [Loarte, 2016-2].
As mentioned above, significant effort in the scenario development R&D has focused on the highperformance flat-top phase of the plasma scenarios. The dynamic behaviour of e.g., the H-mode
pedestal decay, H-to-L transition or the density behaviour during current ramp-down is, on the other
hand, much less well understood. A greater experimental effort combined with dedicatedsimulations is needed to address these scenario development issues for transient phases (particularly
in high performance terminations), such that reliable and robust termination scenarios can be
developed [de Vries, 2016] for ITER.
5.4.5

Impact of scenario development issues on operation

Scenario development traces the progression of the ITER Research Plan. It is important that at eachstage, the new scenarios that need to be explored can make use of the development andunderstanding gained in previous ITER operational phases. PFPO-1 will dedicate a significant
fraction of experimental time to the assessment and optimization of control schemes that will be thebasis of operation in PFPO-2 and later phases. For example, plasma initiation and start-up in thehydrogen phase will address several fundamental issues for all future operating scenarios andcontrol schemes. Plasma breakdown, possibly assisted with ECRH, and early divertor formation
will be tested. Model validation to enhance the confidence in predictive simulation of the next
operational phase scenarios will accompany the execution of the ITER Research Plan.
The development of the ITER Research Plan will also benefit from validation of plasma scenario
modelling tools on present tokamaks, including models for plasma initiation, control schemes, andmodels to describe the evolution of the temperature, density and current profiles. In particular,radial transport models for electron and ion energy and particles in Ohmic and L-mode plasmas thatwere developed many years ago should be re-examined using the more extensive diagnostic
capabilities of present devices so that they can be used to plan more optimal ITER ramp-up
scenarios that maintain the internal inductance, li, within an acceptable range for control.
The basis for establishing ITER baseline Q = 10 DT operation, the initial major goal of the ITER
Research Plan, is reasonably well established. But the experimental development of advanced (i.e.
hybrid and steady-state) scenarios is more challenging, as they require operation at the limits of the
hardware capabilities of any given tokamak where they are explored (i.e. operation at high power
for a maximum time duration, close to plasma stability limits, etc.). With regards to operation of
ITER in regimes with enhanced core confinement, such as hybrid scenarios, but especially thosethat feature ITBs, none of the predictive models for such regimes are as yet in a position to makereliable projections. For the global scaling approach, the limitation may be intrinsic, in that thedevelopment and sustainment of ITBs depends on local plasma parameters (i.e. on detailed plasma
profiles), which are not captured in scalar databases. For the transport models, while progress has
been made in replicating ITB formation and sustainment, further work is required before projections
can be made with confidence of such regimes to ITER. Advanced operation in ITER with enhanced
core confinement is becoming an increasingly realistic and attractive prospect, but a major
experimental emphasis is required to demonstrate that such an advanced operation is compatible

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with reactor operating conditions. Substantial advances have been made in improving the physics
content and reliability of transport modelling and simulation codes, but a fully consistent andintegrated (core and edge) predictive capability that can accurately describe all transport channels
remains to be developed. These also need to be coupled with control schemes that can be
implemented on ITER, consistent with the diagnostics, PF, auxiliary heating, and current drive
capabilities.

DISCLAIMER
The views and opinions expressed herein do notnecessarily reflect those of the ITER Organization

299

ITR-18-003

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Final report from the Magnetic Axis Working Group:https://user.iter.org/?uid=SSKMMC.
ITER Divertor Monoblock Shaping Workshop, 13 – 15 September 2016
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BIPT67 FPPC update by R. Hunt (First Plasma Protection Components):
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Local heat loads to be expected on plasma-facing components during theSecond Plasma Phase (“Pre-Fusion Power Operation 1”):
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ITER Research Plan Workshop, 26 - 28 July 2016 - Presentations andMaterial: https://user.iter.org/?uid=TL5ACR.
Project Change Request: PCR-738 - Establishment of the 2016 ITER
Baseline: https://user.iter.org/?uid=TL5BR3.

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ITR-18-003[ITER_D_TMT985, 2016][ITER_D_TP3QQA, 2016][ITER_D_TVUZP2, 2016][ITER_D_TXQ5E8, 2016][ITER_D_TVG7YK, 2017][ITER_D_TTPHKZ, 2017][ITER_D_UFAXKQ, 2017][Izzo, 2013][Jackson, 2010][Jackson, 2011][Jeong, 2015][JET Team, 1993][Joffrin, 2005][Joly, 2017][Jones, 2001][Kallenbach, 2010][Kamada, 1994][Kavin, 2013][Kavin, 2015][Kavin, 2016-1][Kavin, 2016-2][Kavin, 2016-3][Kavin, 2016-4][Kavin, 2016-5]

5 MA H-mode helium scenario in toroidal magnetic field 1.8T (20MW of
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PCS Physics Requirements Document for the Preliminary Design:
https://user.iter.org/?uid=TVUZP2.
ITER ECE operation at 1.8T: an assessment:
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Staged Approach Configuration - Preliminary Functional Description
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MPP-M030 Presentation 3 - Capability of WAVS for detection of faulty
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Appendix A: Use of Deuterium Seeding to Characterize Fuel
Retention during PFPO-2 Phase
The possibility of limited operation in deuterium during the non-active phases, or of ‘deuterium
spiking’ in hydrogen discharges, has been discussed in earlier versions of the ITER Research Plan(in Appendix D of the IRP-v1 [ITER_D_2FB8AC_v1.0, 2008], Appendix C of the IRP-v2.2[ITER_D_2FB8AC_v2.2, 2009] and Appendix C of the IRP-v2.3 [ITER_D_346RPL, 2010]). With
the development of the new baseline schedule, the decision in 2011 to eliminate the CFC-tungsten
first divertor and the improved understanding and modelling of retention processes in a berylliumtungsten environment afforded by the ILW experiments on JET, the justification for an early look at
retention before deuterium operations begin has not diminished. This Appendix therefore takes a
new and more quantitative look at a possible deuterium spiking scenario which would allow early
assessment of fuel retention levels and efficiency of fuel recovery, testing of dedicated retention
diagnostics in advance of nuclear operation and calibration of migration/retention models. Such an
experiment may also provide sufficient D outflux for testing of hydrogenic permeation into TBMs
in support of the assessment of tritium permeation rates during FPO. Detailed calculations of thefeasibility of this latter measurement have yet to be made. The example plasma scenario consideredhere can be used as a basis and modified at a later stage depending on the outcome of suchcalculations.
The case chosen is a hydrogen L-mode which would be executed in PFPO-2 at 7.5 MA/2.65 T
using hydrogen neutral beams and ECRH power. The rationale for choosing L-mode is thefollowing:
1. Lower required heating power provides increased guarantee that the discharge can be
executed reliably and repeatedly (a prerequisite if the scenario is to be used for retention
studies);
2. Higher quantities of deuterium gas injection permit to maximize D/H ratio forretention/permeation studies (SOL less opaque);
3. Lower confinement, reducing fusion reactions and hence lower neutronic activation (DDneutrons) and low triton production;
4. More stable conditions, avoiding difficulties of ELMy H-modes and staying away fromoperational boundaries (disruptions).
The experiment, should it go ahead, would be programmed at the end of PFPO-2, benefitting fromthe fact that stable, long-pulse L-mode discharges should be well established at this point.
Moreover, the proposed discharges are intended to provide sufficient material migration formeaningful beryllium deposits to grow in the divertor under well controlled conditions. If, as
proposed, one or more divertor cassettes can be removed following the experiment for post-mortem
analysis of long-term retention (to complement global gas balance), this will only be possible if theexperiment occurs at the beginning of the pre-nuclear shutdown.
To assess the target discharge, and provide input to further modelling specific to plasma-wallinteractions, analysis has been performed using a crude iterative process deploying 1.5-D transportmodelling with the ASTRA code for the plasma core and the SOLPS-ITER plasma boundary code
whose output is used to provide a complete plasma background solution for subsequent migration
calculations in support of fuel retention predictions.
The ASTRA simulations use the framework described in [Polevoi, 2017], based on the analysis in[Pacher, 2015], in which scalings for key SOL and divertor parameters derived from a database of312

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SOLPS-4.3 simulations targeting the baseline burning plasma scenario are used to provide
boundary conditions for the core transport model: separatrix values of Te, Ti, ne, nHe, nDT, theionization source from edge neutrals into the core plasma consistent with the power and particle
fluxes into the SOL and the SOL gas puff strength required to maintain the divertor peak power flux
density at tolerable levels). ASTRA simulates heat and particle transport in the core according toscaling-based models with prescribed profiles for heat diffusivities (assuming e = i) and theempirical relationship D = C with C = 0.2 (see section 2 of [Polevoi, 2017]). It contains models
for pellet ablation and, using the SOLPS boundary scalings, is able to account for SOL opacity andthus select an appropriate balance of core (pellet) and edge (gas) fuelling to satisfy a given
prescription on divertor neutral pressure for given PSOL.
A problem with the approach is that the boundary parameterization derived from the SOLPS-4.3
database applies to a specific (burning plasma) operating space: H-mode like SOL transport (D,0 =0.3 m2s-1, ,0 = 1.0 m2s-1) with 60 < PSOL < 140 MW (the implicit assumption is that this will be at
Ip = 15 MA). These transport coefficients lead to near SOL power channel widths of q,0 ~ 3-4 mmin the SOLPS simulations (see Appendix C).
Applying these boundary conditions to a core transport calculation for a lower power L-mode atlower Ip will not capture the wider SOL resulting from higher L-mode transport and thus will notcorrectly describe the SOL opacity (neutral penetration and thus the core ionization source). To
compensate approximately for this, an adjustment has been made (see below) to the coefficient
multiplying PSOL in the scalings, representing an effective increase in the upstream power width bya factor fq = (/0)1/2(15/Ip), where  is the value fitted in the scaling-based model used in
ASTRA for the core transport simulation to reproduce the L-mode scaling for the energy
confinement time. The degree to which the low power ASTRA simulation output is consistent canthen be checked by running a single SOLPS-ITER case with the same fuel throughput, pumping
speed, PSOL, split between edge and core fuelling, and using the separatrix values of D and 
computed by ASTRA.
The case chosen for the ASTRA study has the following key parameters:
Ip = 7.5 MA, Bt = 2.65 T (q95 = 3), PIN = 30 MW, fd = 0.375
PIN the total auxiliary power, chosen to remain below the L-H transition threshold for a hydrogenplasma at 2.65 T. It is comprised of PECRH = 20 MW, PHNB = 10 MW, provided by reducing the(hydrogen) beam energy to 500 keV. Factor fd  (Gtot/Spump)0.83(PSOL/fq)-0.52 is the desired‘detachment parameter’, equivalent to the normalized divertor neutral pressure  in (see eqn. (1) in[Pacher, 2017]) and is a measure of the degree of detachment (fd = 1 corresponds to the point at
which the total ion flux at the divertor target falls to 80% of its maximum value as divertor neutralpressure, pn increases). Here, Gtot is the total fuel throughput (including H and D) and Spump is thepumping speed. An existing SOLPS-4.3 simulation at PSOL= 40 MW, Gtot = 96 Pam3s-1, D,0 = 0.3
m2s-1, ,0 = 1.0 m2s-1, pn = 3.1 Pa, Spump = 31 m3s-1 (about 50% of the maximum available onITER) is used to fix fd at the chosen value, representing a divertor plasma far from strong
detachment.
Deuterium is assumed to be provided only as gas from the edge, whilst hydrogen can be puffedfrom the edge and/or introduced into the core using high field side pellet injection. For the chosen
fd, the SOLPS parameterization (modified for L-mode transport) provides ASTRA with the edgegas puff strength required to maintain the desired degree of detachment, n e,sep, Te,sep, Ti,sep and Gsep,
the core ionization source due to the neutral influx at the separatrix. The remainder of Gtot isprovided by HFS pellets at the level required to maintain the core density above the beam shine313

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through limit (~31019 m-3 for 500 keV H0). Two ASTRA code runs have been made with these
input parameters assuming that D constitutes 10% and 100% of the input gas puff.
The most important output parameters from these first stage simulations are summarized below:
nsep = 1.151019 m-3, Te,sep = 66 eV, Ti,sep = 101 eV
PSOL = PIN + P - PRAD = 30 + 3.13 – 1.76 = 31.4 MW
ne,avge = 3.21019 m-3, ne,0 = 3.51019 m-3 (Greenwald fraction = 0.52)
Te,0 = 4.6 keV, Ti,0 = 4.1 keV
Total core ionization source (coming from SOL gas puff) = 14 Pam3s-1
Core fuelling (HFS pellets) = 53 Pam3s-1, edge gas puffing = 30 Pam3s-1 (Gtot = 83 Pam3s-1)
D,sep = 1.0 m2s-1, ,sep = 6.0 m2s-1 (thus fq = 4.9)
Triton production rate = 3.61011 s-1 (10% D in gas puff), 3.91013 s-1 (100% D in gas puff)
Triton burn-up fraction < 2%
nD/nH = 0.014 (10% D in gas puff), nD/nH = 0.134 (100% D in gas puff)
Maximum possible flat-top duration ~100 s
qpk = 0.39 MWm-2 (peak divertor power flux density obtained from the boundary parameterization).
The calculated L-H transition threshold power for these conditions is P LH = 44.4 MW, well above
PSOL. Recent results from a pure hydrogen campaign in the ILW environment [Litaudon, 2017]indicate that the maximum core D concentrations found in this ITER D-spiking study (~13%) may
be marginal with respect to reducing PLH and pushing this L-mode discharge closer to the H-mode
threshold. If this were problematic, operation at higher Bt could be considered to move away again
from the threshold.
At these DD fusion rates, the total tritium production is ~1.810-4 g /100 s pulse for 10% D in theedge gas puff. For 100% D gas puffing, the production rate increases by ~100 due simply to theorder of magnitude increase in the core D concentration ( nD2). A proposed migration/retention
experiment aiming for, say, 20,000 s of reproducible L-mode plasma with maximum D
concentration would thus require 200 such pulses, requiring ~2 weeks of machine time (13pulses/day) and producing at most ~ 3.610-2 g and a total DD neutron production (2.45 MeV) of~ 7.2 1015 (assuming 50% probability for the two DD fusion reaction branches: D+D  T+p andD+D  3He + n).
This tritium production for the proposed 10% D-spiked retention/migration experiment is about
twice the 210-2 g estimated as the maximum quantity which could be produced throughout the
non-active phases using hydrogen plasmas due to the natural concentration (~0.015%) of D in H. It
is orders of magnitude lower than the maximum quantity of tritium which can be discharged
annually without a functioning tritium plant in ITER. The resulting vessel activation needs to beassessed and this has not been performed here. The arguments demonstrating acceptable activationin Tritium production and activation for the hydrogen phase will not apply, since the spiking
experiment would be performed in a short, concentrated period, though the levels will still be
extremely low given the very low neutron production from these low performance L-modes.

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To provide a boundary plasma suitable for material migration studies and to check consistency with
the ASTRA run based on boundary parameterization, a SOLPS-ITER simulation has been
performed using some of the ASTRA output to determine the key input parameters:
PSOL = 31.4 MW, D = 1.0 m2s-1, , = 6.0 m2s-1, edge gas puff = 30 Pam3s-1, outflux to the innercore boundary = 53 Pam3s-1.
The simulation has been run in the first instance with pure H and all-metal walls (but no W
sputtering switched on), though SOLPS-ITER can deal with mixed hydrogenic species. The
standard reference magnetic equilibrium at q95 = 3 (~9 cm outboard midplane separation between
first and second separatrix) has been used to generate a numerical grid, even though for a dedicatedfuture material migration/gas balance experiment a more optimized plasma shape would likely bechosen. An existing converged solution at higher density was used to start the simulation, but withdensity scaled down such that ne,sep ~ 1.01019 m-3, close to the ASTRA output value (see above).
Fluid drifts are deactivated.

Figure A-1: Outer (top) and inner (lower) divertor profiles of Te, Ti, target power flux density (q) andparallel ion flux density for the SOLPS-ITER simulation consistent with the ASTRA code run at ne,sep ~1.01019 m-3, PSOL = 31 MW.

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An example of the code output is shown in Figure A-1, which compiles profiles of Te, Ti, ne, q andparallel ion flux density, ||,i at the inner and outer divertor targets (q is the perpendicular heat flux
density assuming an axisymmetric (unshaped) divertor).
This simulation has the following upstream separatrix parameters:
ne,sep = 1.01019 m-3, Te,sep = 100 eV, Ti,sep = 150 eV,
and it is comforting to note that the upstream power flux width correction to the SOLPS
parameterization in ASTRA, fq ~ 5, corresponds to q,near = fqq,0 ~ 17 mm, to be compared with
q,near ~18 mm obtained from this dedicated SOLPS-ITER code simulation.
Likewise, the calculated SOLPS-ITER core ionization source is 14.9 Pam3s-1, extremely close to the
ASTRA value of 14 Pam3s-1 for the same gas puff and core fuelling. This means that the modified
boundary parameterization for the L-mode case produces a core source consistent with the fullboundary plasma simulation and that the core D concentrations can be trusted.
A total of 3.3 MW is radiated in the divertor, all from neutral hydrogen. The neutral pressure is
2.1 Pa. As shown in Figure A-1, qpk ~1.6 MWm-2 at the outer target, about a factor 4 higher than
predicted by the boundary parameterization. There is a moderate plasma pressure loss in a narrow
region around the strike point (~factor 2) and thus the plasma is not as deeply detached as would be
expected for the low n,sep. Despite the low PSOL, this divertor plasma solution has reasonable qpk andwould likely be suitable for migration studies.
The study described here should in no way be taken as definitive. There is evidently much scope fortuning. Future true integrated modelling would allow scoping and refining, without relying on theapplication of boundary scalings taken out of the range of parameters from which they were
derived. However, the degree of agreement obtained between the key outputs from the dedicated
SOLPS-ITER and ASTRA simulations increases confidence that this case study can act as areasonable basis for concluding that a series of low power D-spiked discharges in the non-active
phase suitable for material migration and gas balance/fuel recovery studies can be realistically
executed without exceeding regulatory limits or requiring a functioning tritium plant.

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Appendix B: L-H Threshold and Toroidal Field Ripple effects
on
H-modes
B.1 L- to H-mode power threshold uncertainties when extrapolating towards ITER
B.1.1 Summary
Due to the large scatter in the ITPA database used for the L-H power threshold prediction for ITER,
its 95% confidence interval gives a wide range of possible values for the L-H transition in ITER:
from 0.54 to 1.86 times the scaling value given by 𝑃𝑡ℎ,𝐷 = 0.049. ̅̅̅
𝑛𝑒 0.72 . 𝐵𝑇 0.8 . 𝑆 0.94 . This
introduces significant uncertainties on the range of plasma conditions over which the H-mode will
be achievable in ITER with the baseline heating schemes. The uncertainties in the scaling reflect the
impact of divertor geometry, input torque, Zeff, etc., on the power required to access the H-mode.
Some of them are expected to lead to lower power threshold in ITER than predicted by the scaling(lower Zeff than C wall machines, lower torque input than NBI-heated present machines, etc.) whileothers could lead to a significantly increased power threshold (divertor geometry impact of thetemperature profile at the target and the associated edge electric field Er, etc.).
The minimum power threshold with respect to the plasma density is expected to range between 20
to 40% of the Greenwald density on the basis of the existing database. A more physics based
scaling has been recently proposed but it is not yet universally accepted.
The impact of the H isotopes and species on the H-mode threshold is based on a range ofexperimental results and has uncertainties, particularly for He plasmas. For H isotopes the power
threshold decreases inversely with the isotopic mass. The impact of He versus D on the power
threshold either leads to higher or unchanged thresholds depending on the experimental devices; so
far these results are unexplained. Thus, for the isotope effect ITER predictions are based on
𝑃𝑡ℎ ∝ 𝑀−1 and for He the power threshold is expected to be between 1 to 1.4 times the D power
threshold.
Until scalings accounting for the numerous specific impacts are available or, conversely, until abetter theoretically based understanding is able to unify the various observations, it is a reasonableapproach to use the ‘n, B, S’ scaling [Martin, 2008]. However, it should always be considered thatthe 95% confidence interval of this scaling is rather large and this can have significant impact on
the Research Plan for ITER.
B.1.2 ITER L-H power threshold estimates
The ITPA 2008 scaling [Martin 2008] is used to extrapolate towards ITER the L to H mode power
threshold such that:
𝑃𝑡ℎ,𝐷 = 0.049 × ̅̅̅
𝑛𝑒 0.72 × 𝐵𝑡 0.8 × 𝑆 0.94 (B.1-1)
𝑆 is the plasma surface, taken to be 683 m2 for ITER plasmas.
The volume line averaged density ̅̅̅𝑛𝑒 (usually quoted as a percentage of the Greenwald density:
𝐼𝑝
𝑛𝐺𝑊 = 𝜋𝑎2 ).
𝐵𝑡 is the toroidal magnetic field on the magnetic axis.

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In H, the power threshold is taken to be such that 𝑃𝑡ℎ,𝐻 = 2 × 𝑃𝑡ℎ,𝐷 .
In He, the power threshold is taken to be such that 𝑃𝑡ℎ,𝐻𝑒 = [1 𝑡𝑜 1.4] × 𝑃𝑡ℎ,𝐷 .
In DT, the power threshold is taken to be such that 𝑃𝑡ℎ,𝐷𝑇 = 0.8 × 𝑃𝑡ℎ,𝐷 .
This leads to the H-mode threshold values displayed in Figure B-1.

Figure B-1: Expected L-H threshold power on the basis for the B.1-1 scaling for H, He and D plasmas in
ITER over a range of plasma currents for q95 = 3 and two values of the plasma density. For He plasmas therange 𝑃𝑡ℎ,𝐻𝑒 = [1 𝑡𝑜 1.4] × 𝑃𝑡ℎ,𝐷 is shown.

B.1.3. Existing ITPA scalings
In [Martin, 2008] the power scaling was derived with 1024 data points. The root mean square(RMS) value of the fit is 30.8%. This means that the 95% confidence interval of the power
threshold is between 54 and 186% of the extrapolated power value. This reflects the very large
deviations in the data used. Even when selecting only JET data, one finds a factor 5 variation of themeasured power threshold with respect to the scaling value, as illustrated on Figure 1 of [Martin,
2008].
The previous ITPA scaling was produced in 2004 [Takizuka, 2004] and it accounted for the impactsof the aspect ratio and effective charge, as well as density, magnetic field and plasma surface. When
reduced to density, B and S this scaling is very similar to that of [Martin, 2008] within a standard
deviation of 0.36:
𝑃𝑡ℎ,𝐷 = 0.06 × ̅̅̅
𝑛𝑒 0.7 × 𝐵𝑇 0.7 × 𝑆 0.9 (B.1-2)
B.1.4 Impact of ion species
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Based on T, DT, D and H experiments at JET, the dependence of the threshold power on the ion
mass number M is given by 𝑃𝑡ℎ ∝ 𝑀−1 [Righi, 1999]. Experiments in H versus D have also been
carried out in other tokamaks to explore further the isotope impact on the power threshold andfound that 𝑃𝑡ℎ ∝ 𝑀−1 provides a good guideline for the isotopic effect, although the precise valueof the multiplier from H to D depends of plasma parameters.
He versus D has also been explored on DIII-D, ASDEX-Upgrade, Alcator C-Mod and JET, most of
the experiments in a C wall environment [Gohil, 2012; McDonald, 2010]. When comparing He withD plasmas, either the power threshold is unchanged (ASDEX-Upgrade), unchanged or somewhat
increased depending on density (JET) or largely increased by up to factors of 3 at some densities(DIII-D and Alcator C-Mod).
Recent experiments from JET show that small amounts of H in D plasmas and He in H plasmas canhave significant effects on H-mode access [Hillesheim, 2016]. Although this was identified alreadyin ASDEX-Upgrade for He plasmas [Ryter, 2013], the new JET results indicate that the issue ismore complex and can be beneficially exploited to expand the range of ITER H-mode operation in
H plasmas.
B.1.5 Impact of density
The [Martin, 2008] scaling exhibits a threshold power increasing with increasing density. In many
experiments, the power threshold is minimized at a given density and then rises up at low density
again. The position of this minimum density is reported to range between 20 to 40% of theGreenwald fraction [Martin, 2008], as shown in Figure B-2.
Recent work on ASDEX-Upgrade [Ryter, 2014] showed that this minimum density is determinedby the role of the edge ion heating in triggering the L-H transition. Based on these results, an
expression for the density providing the minimum threshold L-H power has been derived, which
also describes well the values found in tokamaks of various sizes.
𝑠𝑐𝑎𝑙(1020 𝑚−3 ) ≈ 0.07 𝐼𝑝0.34 𝐵𝑡0.62 𝑎−0.95 (𝑅/𝑎)0.4 (B.1-3)
𝑛𝑒,𝑚𝑖𝑛For ITER, this leads to,𝑠𝑐𝑎𝑙
𝑛𝑒,𝑚𝑖𝑛𝑛𝐺𝑊

𝐼 0.34 𝐵0.62≈ 0.71 𝑝 𝐼 𝑡
𝑝(B.1-4)

For q95 = 3 conditions in ITER, for which Bt (T) = 0.35Ip (MA), this corresponds tone,minscal ~ 0.4nGW. Accessing the H-mode at these densities reduces the additional heating power
required and this is particularly important for the access to H-mode at 1.8T in PFPO-1 when the
installed additional heating power will be in the range of 20-30 MW.
It should be mentioned however, that in JET with the ITER like wall access to H-mode at low
densities exhibits a significant increase, as in the ASDEX-Upgrade experiments, although Te = Ti,
thus questioning the findings in [Ryter, 2014] that the increase of H-mode threshold at low power isdue to the decreased ion power flux due electron-ion thermal decoupling at low densities [Maggi,
2014].
B.1.6 Impact of toroidal field ripple
TF ripple up to 1.1% has been found not to affect the H-mode threshold in JET experiments[Andrew, 2008]. In earlier JET experiments, when switching from 32 to 16 coils, hence with aripple factor reaching 10%, the H-mode was obtained at a slightly higher power [JET Team, 1993].
Conversely, on JT-60U, the power threshold was reduced when increasing the ripple (before319

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ferromagnetic inserts) [Tobita, 1995]. In DIII-D, a TBM mock-up insert coil induced a local 3%
ripple [Gohil, 2011], which did not modify the L-H threshold value nor its dependence on input
torque.
The impact of resonant and off-resonant fields applied for ELM control has been studied on
ASDEX-Upgrade, MAST and DIII-D [Gohil, 2012; Fenstermacher, 2010; Ryter, 2013]. Strong
resonant components in the Resonant Magnetic Perturbation (RMP) spectrum lead to significant
increase in Pth for RMP fields above a critical threshold field (i.e. for dB/Bt above ~310-4),
whereas no clear increase in Pth is observed below this field. For off-resonant fields, there is a muchsmaller increase in Pth with the applied fields.
Therefore, a priori no significant impact of the magnetic field ripple on H-mode access is expectedin ITER, even at the 1.3% expected at 1.8T. On the other hand RMP fields applied for ELM control
are expected to have a large effect on the L-H threshold when optimized for ELM control.
Therefore the application of these fields when accessing the H-mode should be carefully designed
to minimize their effect on H-mode access. Nonetheless, it should be noted that the normalized
magnitude of the field perturbation either by ripple of RMPs is not an appropriate dimensionlessparameter that should be used to extrapolate towards ITER. More modelling on existing
experiments should take place, and the collisionality regimes in which effects are found should becharacterized and compared to the ones expected in ITER, in order to draw physics-based
conclusions on the effects of TF ripple and 3-D fields applied for ELM control on the H-mode
power threshold in ITER.

Figure B-2: Plasma density at which the minimum H-mode threshold power is achieved versus toroidal field
as a fraction of the Greenwald density (from [Martin, 2008]).

B.1.7 Impact of divertor geometry (Vertical Target versus Horizontal Target and X- point
height)
The L-H power threshold has been found to be lower by 20 to 35% with increased divertor closure
in JET with Carbon PFCs [Horton, 1999; Andrew, 2004] and in JT-60U [Fukuda, 2000]. In more
recent Alcator C-Mod experiments, the slot divertor configuration is associated with a lower power
threshold (up to a factor 3) than the vertical target configuration [Ma, 2012].
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In DIII-D a lower X-point height leads to a lower threshold (a factor 2 for a 15 cm variation)[Gohil, 2011]. A similar X-point height impact is also reported for JET with Carbon PFCs (a factor
5 for 25 cm variation) [Andrew, 2008]. More recently in JET with the ITER-like Wall, experiments
to determine the impact of the vertical versus horizontal target divertor geometry on H mode access,have shown that up to a factor 3 reduction of the threshold is obtained with the horizontal target
configuration [Delabie, 2015]. In this later case a qualitative explanation has been proposed byusing EDGE-2D/Eirene modelling that shows that lower temperature gradients on the target plate
are expected and measured in the vertical target case, leading to lower positive E r at the separatrix
and, hence, potentially reduced EB shear [Chankin, 2017]. A reduced EB shear is qualitatively
consistent with an observed higher power threshold [Burrell, 1990].
B.1.8 Metallic wall impact on H-mode access
The power threshold is found to be reduced with metallic walls as observed in JET and ASDEXUpgrade [Maggi, 2012; Neu, 2013]. Several authors have, thus, suggested that the effective charge
could be an important parameter [Takizuka, 2004; Bourdelle, 2014]. JET-ILW and C wall data havebeen shown to be better fitted by the [Takizuka, 2004] scaling, including a Zeff dependence than by
the [Martin, 2008] scaling [Maggi, 2014]. This trend is confirmed by N2 seeding impact on the
power threshold demonstrated in JET [Delabie, 2016].
B.1.9 Torque impact
In DIII-D using balanced NBI, a lower Pth was found for lower torque input [McKee, 2009]. In
ITER, which has a low normalized torque due to the high energy NBI, a lower power threshold
might be observed compared to the value extrapolated from the existing scalings, such as [Martin,
2008], where a large number of data are provided from NBI heated plasmas with a significant
normalized torque.
B.1.10 Needs for further understanding/characterization
Further studies are required to refine the evaluation of the ITER H-mode power threshold:

From the existing ITPA ‘n, B, S’ scalings, the confidence interval at 95% is very wide andleads to large uncertainties of 54% and 186% around the predicted power threshold. This isdue to hidden parameters that have to be investigated further and understood, including:divertor geometry, X-point height, torque, etc. SOL modelling relating the target
temperature to the separatrix radial electric field, as done recently on JET [Chankin, 2017],
has to be applied to other tokamaks. The nature of the turbulence stabilization when entering
the H-mode has to be characterized and its parametric dependences determined as well.
Indeed the competition between turbulent drive and EB shear is the most commonly sharedbasis for L-H transition understanding.

The isotopic effect needs to be further understood in detail. As DT experiments are rare, it isimportant to perform further H versus D experiments, particularly with metallic walls, todevelop the physics understanding that can be used for ITER predictions.



Depending on the device, either both He and D have the same H-mode power threshold or
the power threshold is up to 3 times higher in He. Further experiments, in metallic
environments should be performed. Models should address the issue of He versus D impact
on the power threshold as well as the impact of small amounts of He into H plasmas, as thishas large implications on the ITER Research Plan H-mode studies in the PFPO phases.

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The density impact on the L-H power threshold also has to be investigated. In particular theincrease at low density due to a decoupling between ions and electrons might explain all
experimental observations. Further consolidation should be provided to these ASDEXUpgrade experimental observations.



The impact of the TF ripple and non-axisymmetric fields on the power threshold observed inexisting tokamaks should be modelled and extrapolated to ITER to determine its potential
effect for the collisionalities pedestal conditions expected in ITER plasmas. This could be
carried out by applying stellarator neoclassical transport codes (SFINCS, PERFECT) to
ITER [Paul, 2016] and theoretical work, such as [Garbet, 2010], should be applied to ITER
to predict the Er modification due to these effects.

B.2 Effects of the Toroidal Field ripple on H-modes
B.2.1 The effect of Toroidal Field (TF) ripple on pedestal and confinement
In JET, the TF system was configured to feed different currents to the odd and even set of coils outof 32 TF coils [Saibene, 2008]. In this operation mode, the peak TF ripple δR can be actively varied
by selecting an appropriate differential current between each set of coils. In these experiments, fourlevels of δR = 0.08%, 0.5%, 0.75% and 1% were used. The TF ripple in the JET tokamak, where all
TF coils are fed with an equal current, is normally very small at the nominal separatrix radius at theouter midplane (0.08%). A series of H-mode experiments were conducted at 2.6 MA/2.2 T (q95~3)
varying R within the range above. Figure B-3.a shows the normalized confinement H98 as afunction of R. The upper envelope in this plot, which corresponds to the case of unfuelled plasmas(*ped ~ 0.04), indicates a gradual decrease of H98 with TF ripple, with up to ~ 20% confinement
loss for δR = 1%. Figure B-3.b shows the pedestal pressure as a function of TF ripple in this seriesof experiments, where the high density data are excluded because of the lack of Te measurements.
This shows that the confinement degradation with TF ripple at low density clearly originates from a
reduction in the pedestal pressure. On the other hand, the variation of H98 for high density
discharges (see the lower envelope of the plot in Figure B-3.a) is very small and H98 appears to be
independent of TF ripple at higher density and/or higher pedestal collisionality (*ped > 0.1). These
results are consistent with the observation of a much-reduced density pump-out as the density is
increased, independent of TF ripple, as shown in Figure B-3.c. The most striking effect of TF ripple
is the strong density pump out, particularly for the case when no gas puffing is applied. As gas
fuelling is increased, the steady state density achieved in the discharges becomes more and moresimilar despite of the increased TF ripple. At the highest fuelling rates (corresponding to nped ~ 60 –70% nGR), the difference between an H-mode with 0.08% and 1% TF ripple becomes very small interms of achievable density and temperature as well as the energy confinement.
a)

b)

c)

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Figure B-3: Dependence of a) H98, b) pedestal pressure and c) line averaged electron density on TF ripple
amplitude in the JET ripple scan experiments at 2.6 MA/ 2.2T.

In JT-60U, ferritic inserts (FIs) were installed in the vacuum vessel to reduce the TF ripple in 2005
[Shinohara, 2006]. Similarity experiments between JET and JT-60U with variable TF ripple werecarried out for 1.1 MA/ 2.0T plasmas [Urano, 2011] because the FI correction in JT-60U was
optimized for TF ripple reduction at Bt < 2 T [Shinohara, 2006]. While the increased TF ripple
significantly changed the ELM behaviour as described later, the reduction in the pedestal pressure
with TF ripple was small at 1.1MA for both devices as shown in Figure B-4.a. This may indicatethat the effect of TF ripple could be more significant at lower collisionality (*ped ~ 0.04), which isconsistent with the very small confinement degradation with TF ripple in JET 2.6 MA plasmas at
high density (see Figure B-3.a). This characteristic is also seen in the impact of TF ripple on density
pump-out in JT-60U and JET. Figure B-4.b and Figure B-4.c show ne/nGR as a function of TF ripplefor discharges without gas puffing at low (~1 MA) and high Ip (2.6 MA), respectively.
(c)

(b)

Figure B-4: a) Dependence of pedestal pressure on TF ripple amplitude in JET 1.1 MA, 2.6 MA and
JT-60U 1.1 MA ripple scan experiments. Greenwald density fraction as a function of TF ripple forunfuelled plasmas with b) low (~ 1MA) and c) high Ip (2.6 MA).

A strong density pump out of 20-30% at R of 1% is observed at 2.6 MA in JET whereas the
reduction in density is negligible at ~ 1 MA with high collisionality (*ped > 0.1). In JET, the effectof TF ripple on H-mode characteristics was also examined at the reduced Ip/Bt of 1 MA/ 1T (q95 ~3.6) [Saibene, 2008]. In contrast to the results obtained at higher Ip and Bt, the global and pedestal
parameters of the 1 MA/ 1T series of H-modes do not change significantly for increasing TF ripple
values at high collisionality (*ped ~ 0.1), as shown in Figure B-5. In particular, the plasma stored
energy is only slightly reduced at high ripple with constant input power. Despite this, a significant
effect of TF ripple on edge toroidal rotation is observed; the plasma rotation is reduced as the ripple
is increased, with VTOR approaching 0 at the plasma periphery for δR = 1.0%.
B.2.2 Effect of Toroidal Field ripple on type-I ELMs
JT-60U ripple scan experiments showed that ELMs are more sensitively affected by TF ripple thanconfinement [Aiba, 2011]. Figure B-6 shows the dependence of ELM frequency, ELM energy loss
and edge toroidal rotation frequency on the TF ripple in JT-60U 1.1 MA H-mode plasmas at fixed
Psep ~ 6 MW. By the installation of FIs, the ELM frequency decreased by ~ 30% and the ELM
energy loss increased by a factor of two. The power carried by ELMs increased by ~30% withincreasing ripple for a given loss power through the separatrix Psep, at ~ constant plasma energy
indicating that the inter-ELM transport is reduced with the increased TF ripple. The edge toroidal
rotation, which was in the counter Ip direction because of large TF ripple of ~ 1% before theinstallation of FIs, was increased towards the co-Ip direction when the FIs were introduced.

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Therefore, the significant changes to ELMs in JT-60U when FIs were introduced could be related tothe change in the edge toroidal rotation.

Figure B-5: TF ripple scan H-mode discharges at 1MA/1T in JET.

Figure B-6: Dependence of a) ELM frequency, b) ELM energy loss and c) edge toroidal rotation
frequency on the TF ripple amplitude in JT-60U 1.1MA H-mode plasmas at Psep ~ 6MW.

In the JET ripple scan experiments at 1 MA/1 T, the effect of TF ripple on edge toroidal rotation
was also clearly observed, whereas the change in the pedestal performance was very small[Saibene, 2008]. Figure B-7 shows the dependence of normalized ELM energy loss on edge
collisionality in the JET ripple scan experiments at 2.6MA/2.2T. This plot indicates that, for similaredge collisionality, the size of the ELM energy loss is reduced as the TF ripple increases. This isconsistent with the JT-60U findings where increased edge toroidal rotation in the counter direction
leads to smaller ELMs, while the pedestal parameters are not largely changed [Aiba, 2011].
B.2.3. Implications for ITER 5 MA/ 1.8T operation
From the experimental evidence, the effect of TF ripple on H-mode confinement is not uniform but
seems to depend on the plasma parameters, such as the edge collisionality. The JET ripple scan324

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experiments at 2.6 MA/ 2.2T, which show 20-30% loss of pedestal pressure at R ~1%, were
performed at ped* ~ 0.04 for the unfuelled plasmas, which is a similar collisionality to that
expected for Q =10 15MA/5.3T operation in ITER. On the other hand, the JET ripple scan
experiments at 1 MA/1 T and JET/JT-60U similarity experiments at 1 MA, which show very small
loss of the pedestal pressure and confinement at R ~1%, were performed at relatively high ped* >0.1.

Figure B-7: Dependence of normalized ELM energy loss on edge collisionality in JET ripple scan
experiments at 2.6MA/2.2T.

If we assume that collisionality is the key parameter, we can then use the ITER predictions for thepedestal pressure derived from EPED1 [Polevoi 2015] to evaluate the possible effects of ~ 1%ripple on the pedestal plasmas for 5MA/1.8T plasmas in ITER, by first evaluating theircharacteristics without ripple effects. For q95 ~3 the pedestal pressure is found to scale with pped ~Ip2. For plasmas with constant <ne>/nGW, nped ~ Ip and this leads to ped ~ Ip-1. (i.e. the collisionalityof 5 MA plasmas with the same <ne>/nGW that Q = 10 plasmas be three times larger than that of 15
MA Q=10 plasmas). However, for the expected plasma conditions in ITER 5 MA/1.8T plasmas(see appendix G), we expect nped/nGW|5MA = 0.5 nped/nGW|15MA-Q=10 and this reduces further theplasma collisionality of these plasmas so that finally for the foreseen ITER 5 MA/1.8T H-mode
plasmas ped |5MA ~ ped |15MA-Q =10. Therefore, ripple effects are expected to be similar at 5MA/1.8T to those for collisionless experimental plasmas with ITER-like pedestal collisionality
implying a reduction of the pedestal pressure by 20-30%. The level of ripple in ITER at 1.8T may
be decreased by reducing the minor radius of the plasma and displacing it towards the HFS to
reduce the TF ripple amplitude.
From the viewpoint of ELM energy losses and heat loads, this high level of ripple is expected todecrease the ELM energy losses by at least a factor of 2 at R ~ 1%, on the basis of the JET and JT60U ripple scan experiments. However, these experimental results could be due to the change in theedge toroidal rotation enhanced towards the counter direction by the fast ion losses induced byincreasing TF ripple. If this is the case the ELM energy loss reduction with high ripple may notmaterialize in the electron-heated 5MA/1.8T ITER H-mode plasmas.

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Appendix C: Heat Load Management
In the progressive approach to scenario development and preparation for nuclear phase operation,
an important issue is the point in operational space at which the divertor target heat load begins torequire mitigation. This Appendix attempts to provide some approximate guidelines.
When discussing heat load management, it is in general understood that this refers to steady state
power loads. In the case of ELMy H-mode plasmas, this steady load is an average of the inter-ELM
and ELM phases. The approach adopted at the ITER Organization during the divertor and first wall
design phase has been to estimate the expected power loading using static (i.e. time independent)
runs with the SOLPS suite (version 4.3, coupling the B2 fluid code to the Eirene kinetic Monte
Carlo neutral code, no fluid drifts). The usual strategy over the years (documented in a number ofpublications, see e.g. [Kukushkin, 2003]) has been to provide a large number of independent
SOLPS code runs, in which key engineering parameters are varied (e.g. power into the SOL (P SOL),
fuelling throughput, pumping speed), and from which scaling relationships can be constructed foruse as boundary conditions for core plasma models (e.g. separatrix averaged values of electron andion temperatures, electron densities, etc.).
The original SOLPS-4.3 database constituted mostly burning plasma simulations with CFC divertor
targets, but this has now been replaced by a new dataset in which carbon is eliminated in favour of
all-metal walls, with the caveat that tungsten impurities (W) themselves are not simulated. To do sowould be extremely costly in a computational sense and would bring little additional information to
the question of power loading assessment since, if W is present in any quantity sufficient to modify
the divertor power balance, then the quantities are too high to be considered trace and thesimulation would not correspond to a realistic operational scenario.
Part of the database for 60 < PSOL < 120 MW was published in [Pacher, 2015] with emphasis on
burning plasma operation. Several new data points now exist for PSOL < 60 MW. Each point in the
database corresponds to a fully converged SOLPS run (usually requiring several weeks or monthsof clock time), with careful attention paid to particle balance and using the full Eirene capability.
Simulations have been performed with both nitrogen (N) and neon (Ne) seeding which, at thehighest input powers, is required if peak divertor power flux densities are to remain withintechnological limits (qpk ≤ 10 MWm-2).
Given the lack of a realistic method for modelling ELMs in SOLPS and the fact that time dependentcode runs are extremely costly (and moreover cannot deploy a physically correct model of theELM), the simulations assume that the chosen PSOL comprises the power conducted/convected into
the SOL both during and in-between ELMs. This is thought to be a reasonable approach for smaller(e.g. mitigated) ELMs for which the inter-ELM and ELM footprints on the targets are expected tobe similar.
A fundamental issue when extrapolating to ITER is the choice of cross-field transport which, in thecase of heat transport, sets the near-SOL heat flux width, q,near and thus the peak target power flux
density. In recent years, the deployment of high quality infra-red monitoring of divertor targets has
allowed an improved determination of the H-mode inter-ELM q,near, which is now being found incurrent devices to depend essentially only on 1/Ip (or 1/Bpol) [Eich, 2013]. This is preserved at least
up to the largest device (JET) currently in operation and such a scaling is well matched by a theory
in which cross field motion in the separatrix vicinity is assumed dominated by (neo)classicalmagnetic drifts, giving q,near ~2apol/R [Goldston, 2012]. For the 15 MA ITER H-mode reference
burning plasma scenario, this theory predicts an extremely narrow heat flux channel, q,near ~ 1 mm.
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Such narrow widths would in practice be difficult to handle in terms of target power loads.
Simulations with SOLPS-4.3, in which heat and particle transport is reduced to very low values toreproduce ~1 mm near SOL widths, find that the additional divertor dissipation required to bring qpk
to manageable levels results in extremely high, and perhaps unsustainable, upstream separatrix
densities [Kukushkin, 2013]. Very recent calculations using the XGC1 edge gyrokinetic code arefinding, however, that at high current at the ITER scale, it is filamentary (‘blobby’) electron
turbulence which dominates over ion magnetic drift, resulting in a much wider q,near ~ 5-6 mm[Chang, 2017]. It is not currently known if, and at which point in operating space (namely Ip), there
would be a transition from one type of transport to the next, so that a prescription for cross-fieldheat transport in boundary simulations for ITER is uncertain. Moreover, the validity of the XGC1
simulations needs to be further explored.

Figure C-1: Left: dependence on sub-divertor neutral pressure of the peak target power flux density for avariety of operating scenarios from non-active operation at low PSOL in H and He to the highest performance
at QDT = 10. Right: the corresponding upstream separatrix densities.

All SOLPS-4.3 simulations performed thus far and which have been used to build scaling
relationships for core modelling have concentrated on the baseline equilibrium at q95 =3(Ip = 15 MA/Bt = 5.3 T) for fixed values of cross-field transport coefficients for particles(D = 0.3 m2s-1) and heat ( = 1.0 m2s-1). This yields q,near ~ 3-4 mm in the simulations, thus afactor ~3 larger than the values predicted by the ion orbit drift model at 15 MA, but lower than the
more optimistic XGC1 simulation results. The code runs performed at lower PSOL, for H and He
plasma, designed to provide some guidance for non-active phase operation, use the same baseline
equilibrium at q95 = 3 (thus appropriate to Ip[MA]/Bt[T] = 15/5.3, 7.5/2.65, 5/1.8) and assume thesame cross-field transport as for the burning plasma. No attempt has been made to vary the radialtransport inside the separatrix so that these simulations do not include any form of
H-mode transport barrier.
A subset of the ITER SOLPS database, is shown in Figure C-1 (left), where qpk is given as afunction of sub-divertor neutral pressure, pn, a key control parameter for particle exhaust and astrong indicator of the level of divertor detachment. Here, qpk may be either on the inner or outertargets, but is usually found on the outer. Results are included for H, He and DT operation with
PSOL = 20 – 100 MW. The right-hand plot in the figure gives the values of upstream separatrix

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density, nsep in the simulations, demonstrating that the points at lowest pn correspond tonsep ~ 51018 m-3.
To put the fixed transport assumption into context, the ion drift based theory in [Goldston, 2012]referred to earlier gives q,near ~ 2-3 mm for a helium H-mode plasma at 7.5 MA/2.65 T and q,near ~2 mm for a hydrogen H-mode at 5.0 MA/1.8 T, so in both cases not too far from the heat flux
channel width produced by SOLPS for the specified fixed cross-field transport strength. Since
SOLPS is a 2-D code producing only purely axisymmetric heat fluxes, the values of qpk in Figure C1 have been scaled up by a factor 1.4 to account for the global divertor target tilting and front
surface monoblock shaping used at ITER to ensure avoidance of power overload on leading edges[Pitts, 2017], which reduces the divertor effective area for power handling.

Figure C-2: Outer target profiles of ne (upper) and Te (lower) for the H and He scenarios in Figure C-2 atlowest pn.

This data demonstrates that for all but the lowest PSOL (20 MW), it is possible to find boundary
solutions in which qpk can exceed the nominal limit of 10 MWm-2, sometimes significantly. Thisoccurs in general for sub-divertor neutral pressures, pn ≤ 1 Pa. At the highest PSOL, corresponding to
DT operation, higher pn will be required as well as extrinsic seeding (Ne and N have been used forthe high power scans in Figure C-1) to remain at tolerable qpk. Note, however, that it is generally
found that pn ≥ 10 Pa corresponds in the simulations to a state of rather deep detachment, which
may be considered too close to operational boundaries (e.g. H-L back transitions, disruptions). The
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inclusion of shaping, which increases the field line attack angles, reduces the operating space withrespect to acceptable target heat loads.
With regard to the operating points at low pn when target power handling becomes an issue, they
are associated in the simulations with a much lower degree of detachment in the strike point region,
corresponding to high strike point plasma temperatures. This is illustrated in the steady state target
profiles compiled in Figure C-2 for the PFPO operating conditions (namely only H and He plasma)in Figure C-1, at fixed pn ~0.7 Pa. At these low neutral pressures, for any PSOL > 20 MW, strike
point values of Te are in the many tens of eV range, corresponding to incoming ion energies close toor exceeding the tungsten sputtering yield of fuel ions (H, D, He, T), even without accounting forthe higher yields in the case of impurity sputtering (e.g. Be or seeded species). Higher T e would
also, however, increase the prompt redeposition fraction and ELMs, if present, would likely
dominate the W net source [Dux, 2017]. Additional simulations are required to address these points
to estimate the net source in each case and then to assess the efficiency with which the released W
may transport to the core plasma.

Figure C-3: Upper: the data of Figure C-1 transformed in terms of monoblock top surface temperature
values. The purple hashed region corresponds to operating points above the W recrystallization
temperature. Lower: estimate of the allowed operating time to recrystallize to a depth of 1 mm from the
top surface.

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The qpk data in Figure C-1 can be straightforwardly associated with a divertor monoblock front
surface temperature value using finite element thermal analysis taking into account the toroidal
bevelling of the front surface and the appropriate water cooling parameters. This yields the data in
Figure C-3 (upper), where the computed temperature is taken at the toroidal centre of the monoblockand does not, therefore, account for toroidal gap edge heating, a phenomenon which is expected tooccur as a result of the castellated nature of the divertor target (namely the use of discretemonoblocks) and the particularities of the magnetic geometry, together with the presence finite ion
orbits [Gunn, 2017].
Surface temperature is the key parameter determining the proximity to W recrystallization, longknown to be associated with changes in mechanical properties such as a decrease in hardness andstrength. Such modifications strongly affect the thermal shock resistance of W and may favour theappearance of deep macrocracks in the centre of the monoblocks (due to the inherent thermal
gradients provoked by the geometry of the monoblock) and/or the development of surface crack
networks during ELMs [De Temmerman, 2017]. The hashed region in Figure C-3 (upper)
represents the approximate operating domain over which the surface recrystallization temperature(Trecrys) would be exceeded for the ITER divertor W grade.
In fact, recrystallization is a function of both material temperature and time. Operation above Trecrys
may thus be permitted, but for strongly decreasing shorter times as the surface temperature
increases (the recrystallization fraction depends exponentially on time). The allowable operating
space may be characterized through the use of recrystallization kinetics [De Temmerman, 2017].
This has been applied to the temperature data in Figure C-3 (upper) to produce the curves in Figure
C-3 (lower), which provide the operating time allowed before recrystallization to a depth of 1 mm
into the monoblock surface occurs as a function of divertor neutral pressure.
As mentioned earlier in connection with qpk, what can be seen immediately in the plots of Figure C3 is that steady state operation at the lowest neutral pressures is to be avoided once heating powers
in the PFPO phases permit PSOL ≥ 40 MW. Such low pressures correspond to low upstream
separatrix densities (Figure C-1: lower), but these are not inconsistent with values that may be
expected for operation at lower Ip. Although simulations have not yet been performed to quantify
this, it is also to be expected that extrinsic seeding will be much less efficient for low p n operation(high temperatures, lower densities) and thus this may not be an option to reduce qpk in these
conditions. The only approach will be to seek operating points at higher pn (provided this iscompatible with the required nsep), in which case Figures Figure C-1 and Figure C-3 demonstrate thatsteady H-mode operation for essentially any accumulated plasma time will be possible for any PSOL
before nuclear operation. Note that once the ‘threshold’ value (to stay below Trecrys) of pn isachieved for He plasmas, qpk reduces rapidly with pn due to the beneficial additional divertor
radiation obtained in helium compared to operation in hydrogen.
The analysis described above has been confined to steady state operation, using time independent
SOLPS simulations which do not separate the ELM and inter-ELM components in the powertransported into the SOL plasma. Separate operational limits will be imposed when the ELM
transient is accounted for and these are discussed below.
A clear limit on the allowed ELM energy density is given by the requirement to avoid W melting.
This has recently been assessed in a detailed study [Gunn, 2017], which combines 3-D ion orbit
calculations with monoblock thermal analysis to estimate the parallel ELM energy flux densities
which would lead to both monoblock top surface melting and toroidal gap edge melting at the outertarget for the 3 combinations of Ip, Bt at fixed q95 = 3. The analysis properly accounts for: the ion
species and charge, the nominal ELM rise times for H-mode pedestal parameters characteristic of330

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the various scenarios, the baseline monoblock geometry (including the assumption of worst casemisalignments and manufacturing tolerances) and the variation of stationary (inter-ELM)
monoblock surface temperature. The results of this analysis are shown in Figure C-4, where thethresholds for toroidal gap and full surface melting are shown for the three Ip/Bt operating points
15/5.3, 7.5/ 2.65, 5/1.8. Figure C-4 also includes the recent multi-machine ELM parallel energy
density scaling [Eich, 2017] applied to ITER, for comparison (the shaded region encompasses theuncertainty boundaries in the scaling). If this scaling is correct, H-mode operation on ITER at 5 MA
is outside the boundaries for any monoblock melting. At 7.5 MA, toroidal gap edge melting isalready an issue and full surface melting is still to be expected in 15 MA burning plasmas.

Figure C-4: Threshold on outer divertor target parallel ELM energy density as a function of Ip for fullsurface and toroidal gap edge melting. The yellow shaded area corresponds to the recent ELM energy
density scaling for uncontrolled type-I ELMs (figure modified from [Gunn, 2017]).

These ion-orbit calculations have not yet been performed for intermediate combinations of Ip and
Bt. The analysis in [Polevoi, 2017], whilst not including finite ion orbit effects, does attempt tocombine the SOLPS boundary scalings mentioned at the beginning of this Appendix with the ELM
parallel energy density scaling of Figure C-4 and scalings of pedestal pressure limits based on
predictions from stability codes to estimate the point at which ELM-induced monoblock top surface
melting becomes an issue on toroidally bevelled monoblocks for the full range of Ip and Bt, PSOL
and nsep. The conclusions are slightly more pessimistic than those reached from Figure C-4 for the
7.5 MA/2.65 T case, showing that for the most pessimistic boundary of the ELM parallel energy
density scaling, full surface melting could be achieved for Ip ≥ 5.5 MA. At full toroidal field of 5.3
T this reduces to Ip ≥ 3 MA, though it is unlikely that the large ELMs obtained in this case (WELM~10 MJ) would be compatible with sustained H-mode operation (high ELM energy densities on thefirst wall and large W influxes due to low ELM frequencies).
Ion orbit calculations are being performed to assess if further monoblock shaping can be applied toprevent ELM heat fluxes from reaching toroidal gaps (see presentations at the ITER divertor
monoblock shaping workshop [ITER_D_ TFJJNH, 2016]), since toroidal gap edge melting will
always occur before full surface melting (Figure C-4). Electron beam loading experiments are alsoplanned to study the consequences of the very localized toroidal gap melting on the monoblocks
performance.

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Even if melting is completely avoided, ELMs will also contribute to W recrystallization since,
although the ELM-averaged monoblock temperature may not exceed Trecrys, the ELMs will drive thesurface over Trecrys for a very short time (< 1 ms). Assessing the cumulative effect of many ELMs interms of recrystallization kinetics as performed for the steady state (Figure C-3 (lower)) is
underway and is a function both of ELM amplitude, ELM frequency and the baseline inter-ELM
monoblock temperature.
Finally, it should be stated that the transient energy density threshold for the onset of surface
damage (roughening, development of small crack networks) can be extremely low when a large
numbers of events are involved (as for example in the case of high frequency mitigated ELMs).
Electron beam studies of repetitive fast transients (duration ~0.5 ms) with simultaneous steady stateloading have found that for event numbers exceeding ~105, ELM energy densities should notexceed ~0.13 MJm-2 on the W surface [Loewenhoff, 2011]. This corresponds to ~2 MJm-2 parallelto the magnetic field on ITER and is comparable to, but still lower, than the parallel energy density
required to avoid toroidal gap edge melting in Figure C-4.

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Appendix D: Reference Pulse
A reference or standard discharge is run regularly, and in some cases at the start of every
operational day, in many tokamaks, often motivated primarily to assess wall conditions and theirevolution over a campaign [Ryter, 2002; Pitts, 2007; West, 2009]. Such a pulse can also be used to
track variations in the L-H power threshold, the discharge impurity content and to provide a usefultarget for diagnostic calibrations, for example by providing information on the degradation of firstmirror reflectivity. The plasma behaviour during this reference pulse can help optimize the machine
performance by providing indications on the potential need for conditioning. In Tore Supra or JET,
reference discharges have also been used to monitor the thermal response of the plasma-facing
components and provide information on the presence and evolution of co-deposits [Corre, 2004;
Chatelier, 2007] and variations in component surface emissivity (this is of interest depending onwhether two-colour capability is eventually provided for the ITER IR camera systems or not). Morerecently, a dedicated monitoring discharge has been developed at JET to document wall conditions,but also the long-term evolution of erosion fluxes, material migration, material mixing anddeposition in remote areas [Brezinsek, 2013-2; Krieger, 2013].
In a fully actively cooled device such as ITER, particularly one in which stationary power flux
densities are expected to routinely attain PFC heat handling limits, the regular execution of areference discharge is particularly appropriate. Signs of thermal fatigue, for example due todegradation of bond interfaces, can be identified by the relative overheating of components subject
to the same input power flux density. The thermographic systems on ITER have been designed tocover ~90% of the first wall and divertor surfaces and will be capable, depending on the location in
the field of view, of detecting temperature excursions with sufficient accuracy to determine
anomalous behaviour (see, e.g., [ITER_D_TTPHKZ, 2017]) of sub-units of plasma facing
components. In very critical areas, such as divertor strike point regions, this may allow pre-emptive
action to be taken to ensure that degrading parts of plasma facing components do not continue to beoverloaded leading to early failure.
As mentioned above, the gradual degradation of mirror reflectivity is a major concern for ITER
diagnostics. A substantial R&D program has been underway for several years throughput the ITER
Members’ institutions to develop mirror cleaning techniques and shutter systems are foreseen forseveral diagnostics (see [ITER_D_QNDHTV, 2015]), but a periodically executed reference
discharge will be required to track the performance of diagnostics relying on first mirrors.
To maximize the usefulness of the reference pulse, it should include several phases with different
levels of heating power and different confinement modes. This approach has the advantage that thereference pulse can evolve as the machine capabilities evolve while data from the early phases ofthe discharge can be compared throughout the entire machine operations. The discharge could
comprise an L-mode phase (including limiter operation) and two H-mode phases with different
levels of heating power (to vary the wall ion flux and power to the divertor). Some variation of thedivertor strike-point position could be included. The details of the scenario would need to be
developed, but since the discharge needs to be run already during PFPO-1, it is likely that it should
be run at half current/field (7.5 MA/2.65 T) at least for the non-active phases and partly into FPO.
Once high current operation is routinely established, the pulse would likely be executed uniquely at
the nominal Bt = 5.3 T. The frequency of execution will be decided once operations begin.

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Appendix E: Non-inductive Scenarios with/without LHCD
The achievement of Q = 5 in very long pulse, fully non-inductive plasmas is an ultimate goal of the
ITER project, and it may take a considerable period of experimentation and a variety of upgrades toachieve that goal. Among the potential upgrades of various systems, a lower hybrid (LH) heating
and current drive system has been recently considered as a less favourable option due to thetechnical challenges anticipated in its installation and the lack of research and design activity in thefusion community compared to that being conducted in support of the electron and ion cyclotron
wave heating and neutral beam injection systems. In this Appendix, candidate ITER steady-state
operation scenarios previously reported in several key publications are reviewed and an assessmentthat also includes scenarios more recently developed is presented addressing a specific question:
does the inclusion of an LHCD upgrade in the ITER H&CD capability confer particular advantagesin achieving fully non-inductive operation with Q = 5 in ITER?

E.1 ITER steady-state operation scenarios reported in previous publications
Litaudon has performed a scan of the fusion power multiplication factor [Litaudon, 2006] using the
0.5-D METIS code [Artaud, 2008] at a wide range of the plasma confinement enhancement factor(H98) and density profile peaking factor (ne0/<ne>). In his work, 73 MW of total auxiliary heating
power was assumed including 20 MW LHCD. The flat-top plasma current was allowed to vary withzero loop voltage applied after the start of the flat-top (SOF), and the flat-top phase was extended toachieve stationary non-inductive operation.

Figure E-1: ITER zero loop voltage operation(Vloop = 0): fusion gain, QDT, versus the confinement
enhancement HIPB98(y,2) factor assuming ITER
geometric parameters and with 73 MW applied power(33 MW NBI, 20 MW ICRF and 20 MW LHCD).
Reprinted from [Litaudon, 2006].

334

The results from this scan are shown in Figure
E-1 (reprinted from [Litaudon, 2006]). The
fusion power multiplication factor rapidly
increased along with H98 and density profile
peaking factor. The target fusion gain (Q = 5)
was achieved at H98 = 1.6 with a density
peaking factor of about 1.3, whereas it was
only accessible with a high density peaking
factor (>1.5) at H98 = 1.4. Note that the noninductive scenarios obtained at low H98 and Q
would have a high Greenwald density fraction(above 1.0) as the flat-top plasma current was
reduced while the average density was set to
6.7×1019 m-3. This scan does not provide
information on the operation space achievablewith various ITER H&CD upgrade options.
However, the applied modelling approach isuseful for the assessment of various noninductive scenarios with different confinement
enhancement and H&CD upgrade options.
Therefore, an additional modelling analysis
performed using the METIS code is describedlater in this Appendix to address the specificquestion on the achievement of Q = 5 fully
non-inductive ITER operation and H&CD
upgrade options.

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Various assessments of ITER steady-state operation scenarios developed using different tools andoperation assumptions have been previously reported in the literature. In Kessel’s study [Kessel,
2007], several ITER steady-state scenarios developed using TOPICS, TSC/TRANSP, CRONOS
and ONETWO were compared. In these scenarios with a large variation of the H&CD mix, the flattop plasma currents were assumed to be 8-9 MA. The non-inductively driven current fraction of
about 72% was achieved in the CRONOS scenario by minimizing the use of NBI power, whereas
fully non-inductive operation was achieved in the other scenarios. In the TSC/TRANSP and
CRONOS scenarios, the fusion gain of Q = 5 was achieved with LHCD and high plasma
confinement enhancement (H98 > 1.6). In the ONETWO scenario, Q = 5.7 was achieved atrelatively lower confinement enhancement (H98 = 1.42), assuming, however, a high pedestal
temperature of about 8 keV and a high density of 1.2 times the Greenwald density. In another
CRONOS scenario reported by Garcia [Garcia, 2008], Q = 6.5 was achieved at H98 = 1.7 with only
wave heating schemes including 13 MW LHCD. Wagner studied the various H&CD mixes andcurrent drive efficiencies which are necessary for fully non-inductive steady-state operation[Wagner, 2010]. As an optimized case, 27 MW ECRH, 33MW NBI and 20MW ICRF power wereselected and the confinement enhancement factor was varied from 1.37 to 1.78. In this scan,
Q = 4.4-5.2 was achieved with fNI = 0.82 - 0.90.

Figure E-2: Fusion gain (Q) versus the confinement
enhancement factor (H98) collected from the steadystate scenarios reported in various publications. Filledmarkers represent non-inductive (fNI > 0.95) scenariosand open markers represent scenarios with fNI < 0.95.
The point indicated by ‘1’ represents the ONETWO
scenario. The two points indicated by ‘2’ represent
CRONOS scenarios (fNI < 1.0). The point indicated by‘3’ represents the TOPICS scenario. Various HCD
mixes are categorized as follows: Baseline HCD (orHCD0): within (33 MW NBI + 20 MW ECRH +20 MW ICRF), HCD1: Baseline HCD + 13-25 MW
LHCD, HCD2: Baseline HCD + 2-20 MW ECRH,
HCD3: 53-73 MW ECRH.

Murakami has firstly benchmarked six
different scenario modelling tools, TOPICS,
TRANSP, CRONOS, ASTRAi, ASTRAk and
FASTRAN, by applying the ITER baseline
H&CD mix (33 MW NBI, 20 MW ECRH and
20 MW ICRF) to an ITER steady-state
scenario (Ip = 8 MA) [Murakami, 2011]. The
fusion gains obtained from various noninductive operation scenarios (fNI ~ 1.0) were
3-4, but with a relatively wide range of theplasma confinement enhancement (H98 = 1.31.5). An optimization of the fusion gain was
attempted by reducing the ICRF power to
5 MW, and a fully non-inductive scenario
with Q ~ 4.3 was achieved at H98 = 1.53.
Another scan on the achievable fusion gain atdifferent flat-top plasma current was also
performed by allowing fNI < 1.0. A scenario
with Q = 5 was achieved at Ip ~ 9.5 MA(fNI ~ 0.8), however fully non-inductive
scenario was only accessible at Ip ~ 8 MA andwith reduced fusion gain (Q = 3.5).

Poli has investigated fully non-inductive
steady-state ITER operation with various
H&CD mixes using TRANSP [Poli, 2012]. In
her work, the desired fusion gain (Q = 5) was
achieved at H98 = 1.58 - 1.65 with several H&CD mixes including 20-40 MW of LHCD power.
Kim has studied various ITER steady-state operation scenarios using CORSICA [Kim, 2013]. A
scan of the fusion gain at constant plasma current (Ip = 9 MA) was performed by varying theassumed plasmas confinement enhancement. In this scan with 16.5 MW NBI, 20 MW ECRH and
335

ITR-18-003

20 MW LHCD, Q = 3.1 - 5.1 were achieved at H98 = 1.23 - 1.60 and fNI = 0.67 - 0.87. A fully noninductive scenario developed assuming H98 ~ 1.6, Q = 4.2 was achieved without including LHCD
in the H&CD mix (33 MW NBI and 40 MW ECRH). Polevoi has developed a new approach to
explore the operation space for ITER steady-state operation [Polevoi, 2010]. In his study, the fusion
gain was constrained (Q = 5) and the required confinement enhancement was calculated. A fully
non-inductive operation scenario with 22 MW ECRH and 33 MW NBI required a value of H98 of
1.66.
Table E-1 – ITER steady-state scenarios reported in various publications(Power is in MW and the plasma current is in MA)
AuthorKesselKesselKesselKesselGarciaWagnerWagnerWagnerWagnerMurakami
Murakami
Murakami
Murakami
Murakami
Murakami
Murakami
Murakami
Murakami
Poli
Poli
Poli
Poli
Poli
Poli
Kim
Kim
Kim
Kim
Kim
Kim
Polevoi

Code
TOPICS3
TSC/TRANSPCRONOS2ONEWTWO1CRONOS2case Acase Bcase Ccase D
FASTRANTOPICS
TRANSP
CRONOSASTRAiASTRAkFASTRAN
FASTRAN
FASTRAN
TRANSP
TRANSP
TRANSP
TRANSP
TRANSP
TRANSPCORSICACORSICACORSICACORSICACORSICACORSICA
ASTRA

PNB
33
16.5
5.9
33
0
33
0
0
33
33
33
33
33
33
33
33
33
33
33
33
33
33
33
33
16.5
16.5
16.5
16.5
24.75
33
33

PEC
40
0
0
20
20
20
53
73
40
20
20
20
20
20
20
20
20
20
20
20
40
40
20
20
20
20
20
20
20
40
22

PIC
17
20
16.3
20
20
20
0
0
0
20
20
20
20
20
20
20
10
5
20
20
20
20
0
0
0
0
0
0
0
0
0

PLH
0
25
20
0
13
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
0
20
20
20
20
20
20
20
0
0

H98
1.86
1.63
1.60
1.42
1.70
1.37
1.37
1.37
1.37
1.49
1.48
1.43
1.30
1.36
1.50
1.49
1.50
1.53
1.57
1.64
1.52
1.61
1.63
1.64
1.23
1.34
1.46
1.60
1.59
1.61
1.66

Ip
9
8
8.4
9
8
9
9
9
9
8
8
8
8
8
8
8
8
8
7
7.4
9
9
8.85
8.85
9
9
9
9
9
9
9

Q
3.5
5.1
6.9
5.7
6.5
4.4
4.45
4.3
4.38
3.31
3.32
3.31
3.8
3.34
3.11
3.36
3.38
4.25
2.4
3.6
3
3.8
4.3
5
3.06
3.64
4.32
5.14
4.59
4.24
5

fGW
<1.0
<1.0
<1.0
1.18
0.90
0.97
0.97
0.97
0.97
0.85
0.85
0.85
0.85
0.85
0.85
0.85
0.85
0.85
1.00
0.99
0.84
0.96
0.95
1.05
0.80
0.80
0.80
0.80
0.80
0.80
0.90

fNI
0.98
0.98
0.72
1.01
0.90
0.82
0.72
0.71
0.83
1.06
1.03
0.99
1.06
1.04
1.05
1.02
1.01
1.01
1.00
1.00
0.98
0.95
1.00
1.00
0.67
0.72
0.79
0.87
1.00
0.99
1.00

HCD
2
1
1
0
1
0
3
3
2
0
0
0
0
0
0
0
0
0
0
0
2
2
1
1
1
1
1
1
1
2
2

The ITER steady-state scenarios reviewed in this Appendix are summarized in Table E-1 and also
shown in Figure E-2. The point indicated by ‘1’ represents the ONETWO scenario in which theplasma density was assumed to be quite high (fGW ~ 1.2). The two points indicated by ‘2’ represent
the CRONOS scenarios in which NB power was significantly reduced (<6 MW) and inductive
current was provided to increase the fusion gain (therefore allowing fNI < 1.0). The point indicatedby ‘3’ represents the TOPICS scenario in which H98 is increased to a very high value of 1.86. The
336

ITR-18-003

other steady-state scenarios were inside the region (H98 < 1.66 and Q < 5.1). By comparing the
scenarios reported in these publications, several relevant observations can be drawn. Firstly, aplasma confinement enhancement (H98) of above 1.6 is required to achieve Q = 5 operation (fNI ≤ 1)with the ITER baseline H&CD systems (33 MW NBI, 20 MW ECRH and 20 MW ICRF), if theflat-top density is lower than the Greenwald density and density peaking is limited to 1.5. Secondly,
if H98 = 1.6 is achievable, fully non-inductive Q ~ 5 operation may be accessible with several
H&CD upgrade options including additional ECRH and/or LHCD power. Lastly, one of the H&CD
upgrade options, adding the 3rd HNB, was not explored in the previous studies.

E.2 METIS analysis of fully non-inductive ITER operation scenarios (fNI = 1.0)
Based on the review of the previously developed ITER steady-state scenarios, an additional analysis
has been performed to study the issues related to fully non-inductive ITER steady-state operationand the H&CD system upgrades. Firstly, potential H&CD upgrade options for achieving Q = 5 fully
non-inductive steady-state operations were investigated. The confinement enhancement factor of
1.6 was assumed to allow the plasma to access to Q = 5 at fully non-inductive operation. Secondly,
the confinement enhancement factor has been reduced to 1.2 to examine how plasma fusion
performance varies across the various H&CD upgrade options at lower values of the plasma
confinement enhancement factor. Lastly, a scan of Q versus H98 in fully non-inductive ITER
operation was compared with the previous scan performed at a constant Ip (fNI < 1.0 is allowed)
using CORSICA. These scenarios are summarized in Table E-2 and also shown in Figure E-3.
E.2.1 Assuming a plasma confinement enhancement factor of H98 = 1.6
To study the achievement of Q = 5 fully non-inductive ITER steady-state operation, a confinement
enhancement factor of 1.6 has been assumed based on the literature review previously undertaken.
Such high confinement enhancement is identified as a necessary condition for accessing fully noninductive steady-state operation with Q = 5. Several METIS scenarios with different H&CD
upgrade options have been developed adopting the approach introduced by Litaudon [Litaudon,
2006]. The plasma density was constrained below the Greenwald density along with the achieved
flat-top plasma current at a stationary state with zero loop voltage. Various H&CD mixes including
several upgrade options are categorized as follows:






Baseline H&CD (or HCD0) : Within 33 MW NBI + 20 MW ECRH + 20 MW ICRF
HCD1 : Baseline H&CD + 20 MW LHCD
HCD2 : Baseline H&CD + 20 MW ECRH
HCD4 : Baseline H&CD + 16.5 MW NBI
HCD5 : Baseline H&CD + 16.5 MW NBI + 20 MW ECRH

Note that 20 MW ICRF option was not included in the H&CD upgrade options (HCD1-HCD5) tooptimize the fusion gain. In the non-inductive scenario with only the Baseline H&CD mix (Case1),
Q ~ 3.7 was achieved with a non-inductively driven plasma current of about 7.5 MA. The fusion
gain was increased to 4.9 (HCD1), 4.1(HCD2), 4.7(HCD4) and 5.3(HCD5) when the H&CD
upgrade options were applied assuming a density profile peaking factor of 1.3. This comparison
shows that a similar increment in fusion gain to that achieved by adding 20 MW LHCD power(HCD1) can be also achieved by adding either (16.5 MW NBI – HCD4) or (16.5MW NBI +20 MW ECRH – HCD5). Replacing 20 MW LHCD (HCD1) by 20MW ECRH (HCD2) seems to beless efficient in optimizing Q as the ECCD efficiency off-axis (ρ = 0.4 - 0.5) achieved in thisanalysis was about half of the LHCD efficiency (2.2×1019 m-2AW-1). Adding a 3rd HNB (HCD4 and337

ITR-18-003

HCD5) appears to be more efficient than adding 20 MW ECRH (HCD2). The maximum variationof the fusion gain (𝛥Q) was about 1.6. The fusion gain can be further increased if higher density
profile peaking (> 1.3) or flat-top density (fGW ~ 1.0) can be maintained for non-inductive steadystate operation. Note that the dependence on the density profile peaking has been already studied[Litaudon, 2006; Poli, 2012; Polevoi 2010] and is also shown in the Figure E-1.
E.2.2 Assuming a reduced plasma confinement enhancement factor of H98 = 1.2
The H&CD upgrade options for fully
non-inductive ITER steady-state
operation were also compared at a
reduced confinement enhancement
condition (H98 = 1.2). Operating theplasma at such relatively modest
confinement enhancement has been
demonstrated in many tokamak
devices and may also be more easily
achievable in ITER. In the noninductive scenario only with theBaseline H&CD mix (Case2) asshown in Table E-2 , Q ~ 1.3 wasachieved with a non-inductively
driven plasma current of about
Figure E-3: Fusion gain (Q) versus the confinement
5.9 MA. With the H&CD upgradeenhancement factor (H ) obtained from the scenarios
options, the fusion gain was slightly developed using METIS 98and CORSICA. Open markers (ofincreased to 1.7 (HCD1), 1.6 (HCD4) HCD1) represent CORSICA scenarios at a constant flat-top
and 1.8 (HCD5, Case12), whereas it plasma current of 9 MA (fNI = 0.67~0.87) and the other markers
remained similar with the ECRH represent fully non-inductive METIS simulations (fNI = 1.0).
system upgrade option (HCD2). The The blue dash-dot line is replotted from the red solid line
maximum variation of the fusion gain (density profile peaking of 1.3) shown in the Figure E-1.
was about 0.5, indicating that the Various H&CD mixes are categorized as follows: Baseline
fusion gain achievable at reduced H&CD (or HCD0): Within 33 MW NBI + 20 MW ECRH +confinement enhancement would be 20 MW ICRF, HCD1: Baseline H&CD + 20 MW LHCD,
less sensitive to the H&CD upgrade HCD2: Baseline H&CD + 20&MW ECRH, HCD4: Baseline
H&CD + 16.5 MW NBI, HCD5: Baseline H&CD + 16.5 MW
options. This comparison also shows
NBI + 20 MW ECRH.
that the fusion gain may approachQ ~ 2 with several H&CD upgrade options if the density profile peaking and flat-top density can befurther optimized.
E.2.3 Steady-state operation at a constant Ip (fNI < 1.0)
A scan of the Q versus H98 in fully non-inductive operation has been additionally performed using
METIS (listed in Table E-2), and has been compared with the previous scan at a constant plasma
current (Ip = 9 MA) performed using CORSICA (listed in Table E-1). These two scans arecompared in Figure E-3. The red dashed line shows the scan at constant plasma current (fNI = 0.67 0.87) and the blue dash-dot line shows the scan in fully non-inductive operation (Ip = 7.9 10.1 MA). The fusion gain in fully non-inductive operation was rapidly increased as theconfinement enhancement factor was increased from 1.2 to 1.6. This was due to the strong
dependence of the energy confinement time (τE) on the plasma current. This comparison shows that338

ITR-18-003

the dependence of the fusion gain on the confinement enhancement factor is stronger for the fully
non-inductive operation (fNI = 1.0), implying that achieving high confinement enhancement(H98 ~ 1.6) would be critical for Q ~ 5 fully non-inductive ITER operation. This comparison also
shows that modest fusion gains (Q = 2-5) may be achievable in long-pulse operation (~3000 s) with
H98 = 1.2 - 1.6, if the flat-top plasma current is constrained for better confinement. Note that
20 MW LHCD (HCD1) was included in the CORSICA simulations, whereas the 3rd HNB and an
additional 20 MW ECRH (HCD5) were included in the METIS simulations.
Table E-2 – ITER steady-state scenarios developed using METIS: a density profile
peaking factor of 1.3 was assumed (power in MW, plasma current in MA)
Case
1
2
3
4
5
6
7
8
9
10
11
12

PNB
33
33
33
33
33
33
49.5
49.5
49.5
49.5
49.5
49.5

PEC
20
20
20
20
40
40
20
20
40
40
40
40

PIC
20
20
0
0
0
0
0
0
0
0
0
0

PLH
0
0
20
20
0
0
0
0
0
0
0
0

PAux
73
73
73
73
73
73
69.5
69.5
89.5
89.5
89.5
89.5

H98
1.6
1.2
1.6
1.2
1.6
1.2
1.6
1.2
1.6
1.5
1.4
1.2

Ip
7.5
5.9
9.1
7.4
8.2
6.6
8.5
7.0
10.1
9.2
8.6
7.9

Q
3.7
1.3
4.9
1.7
4.1
1.2
4.7
1.6
5.3
4.3
3.0
1.8

fGW
0.90
0.85
0.93
0.91
0.95
0.89
0.91
0.84
0.84
0.92
0.90
0.85

fNI
1
1
1
1
1
1
1
1
1
1
1
1

βN
2.8
1.9
2.8
1.8
2.8
1.8
2.9
1.9
3.4
3.0
2.6
2.0

HCD
0
0
1
1
2
2
4
4
5
5
5
5

E.3 Plasma stability analysis
A stability analysis has been performed to compare the resulting plasmas from the various H&CD
upgrade options with or without LHCD. Two METIS simulations shown in Table E-2 , Case3(33 MW NBI + 20 MW EC + 20 MW LH) and Case9 (49.5 MW NB + 40 MW EC), have been
selected for this comparison. These two simulations were reproduced using the 1.5-D CORSICA
code to improve the quality of the scenario simulation taking the ITER H&CD configuration into
account as well as to verify the feasibility of steady-state operation in terms of various PF coil andpower supply limits. The modelling of the plasma edge pedestal has been also improved withincreased radial profile resolution as well as using the EPED1 estimates [Snyder, 2011]. As theplasma temperature was more peaked in the CORSICA simulations (Case3C and Case9C in Table
E-3 ), the Q values were slightly higher than those in the METIS simulations (Case3 and Case9 inTable E-2 ). These two CORSICA simulations were then repeated using ASTRA [Pereverzev,
2002] to further improve them by including the fast ion pressures in the evaluation of theequilibrium and plasma . Finally, two plasma stability analysis codes, KINX [Degtyarev, 1997]
and MISHKA [Mikhailovskii, 1997], have been applied and the plasma MHD stability of the two
scenarios has been studied. As MHD stability issues were identified for Case9C, a variant ofCase9C with reduced EC power (30 MW) and plasma density, Case9Ca, has been also additionallydeveloped to reduce the βN to a value close to that in Case3C. This scenario has been also analyzed
to compare MHD stability of plasmas with similar βN values, but different H&CD mix (Case3C vsCase9Ca). All three cases are summarized in Table E-3 and the KINX plasmas stability analysis issummarized in Table E-4.

339

ITR-18-003

To study the MHD stability analysis of a given scenario, βN has been varied with fixed j|| and scaled
pressure gradient (dp/dψ) profiles, and marginally stable values were identified for various MHD
instabilities. In both Case3C and Case9C, all thecritical MHD modes considered were unstable(βN,limit < βN), although the βN,limit values for the n = 1
external mode were slightly increased when thepresence of the ITER wall shown in Figure E-4 wasconsidered (compared to the no-wall evaluation).
The limit for internal modes was set by the n = 3
mode, and this was particularity unstable in Case9C,
with a relatively low βN,limit of 2.54. It appears thatthe higher core plasma pressure gradient, resulting
from the high NBI and EC power levels, is
responsible for such a low βN,limit for the n = 3
internal mode. On the other hand, for Case9Ca,
which has a plasma beta comparable with that of
Case3C and less peaked core plasma pressure
profiles, the βN,limit for the n = 3 internal mode wascomparable (2.98) to that of Case3C. The n = 1
external mode stability is also improved for both the
no-wall and ITER wall studies for Case9Ca
compared to Case9C, and the n = 1 external modewith the ITER wall becomes marginally stable(βN,limit ~ βN). The higher li of Case9Ca compared tothat of Case3C was identified as beneficial for the
improved n = 1 external mode stability of thisplasma. MISHKA stability analysis performed inparallel for the same plasmas has confirmed the
KINX results.

Figure E-4: The model of the ITER conformal
wall (black line) used in KINX stability
analysis.

Based on these plasma stability analyses and their comparison, the H&CD upgrade option which
includes LHCD appears to be less favourable in terms of plasma MHD stability for the achievementof high-N non-inductive plasmas. As LHCD provides driven current at a far off-axis location in
ITER plasmas, the plasma inductance (li) will be reduced compared to values likely to be producedwith H&CD upgrade options which do not include LHCD. For the suppression of the n = 3 internalmode, which may require either an optimization of local pressure gradient at the low magnetic shear
region or an active control of magnetic shear at the high pressure gradient region, upgrading the
ECCD power is shown to be beneficial. As ECCD current drive is less efficient than that of LHCD
for a given power level, this needs to be compensated by additionally adding a 3rd NBI beamline
when LHCD is removed from the H&CD mix. Therefore, considering the stability of target plasmas
for non-inductive operation having similar βN with and without LHCD, the H&CD upgrade option
including both the 3rd NBI beamline and a total of 40 MW EC appears to be equivalent to, andperhaps even more flexible than, the H&CD upgrade option including 20 MW LHCD. It is also
worth mentioning that the values of βN,limit estimated for the n = 3 internal mode in our studies arelikely to be conservative values, because the presence of the fast particles is expected to havebeneficial effects in improving the plasma stability against internal modes; this is an effect which iswell known and characterized for the n = 1 mode (associated with sawteeth) and that needs to beassessed in detail for the n = 3 mode in these plasmas.
340

ITR-18-003

Table E-3 – ITER steady-state scenarios developed using CORSICA: A density profile
peaking of 1.3 was assumed (power is in MW and the plasma current is in MA)
Case

PNB

PEC

PIC

PLH

PAux

H98

Ip

Q

fGW

fNI

βN

3C
9C
9Ca

33
49.5
49.5

20
40
30

0
0
0

20
0
0

73
89.5
79.5

1.59
1.61
1.61

9.0
10.0
10.0

5.7
6.3
5.4

0.94
0.91
0.80

0.99
0.98
0.98

2.91
3.30
2.94

H&CD
option1
55Table E-4 – βN,limit from KINX stability analysis: βN has been varied with fixed j|| andscaled pressure gradient (dp/dψ) profiles, and marginally stable values wereidentified for various MHD instabilities; βN is recomputed using ASTRA
including the pressure from fast -particles and beam-injected ionx; the βN
and li computed from CORSICA are indicated by (*).
Case3CMHD modes
PNB/PEC/PLH=33/20/20 MW
βN=3.27(2.91*), li=0.61*
n=1 no-wall
2.26
n=1 ITER wall
2.70
n=3 internal
2.96

Case9CPNB/PEC=49.5/40 MW
βN=3.78(3.30*), li=0.75*
3.26
3.62
2.54

Case9CaPNB/PEC=49.5/30 MW
βN=3.34(2.94*), li=0.79*
2.78
3.37
2.98

E.4 Summary and Conclusions
From the literature review on modelling of ITER non-inductive steady-state operation, several
conclusions can be drawn:High plasma confinement enhancement (H98 >1.6) was required to achieve Q = 5 operation(fNI ≤ 1) with the ITER baseline H&CD mix and a reasonable density profile peaking
assumption (~ 1.3).
Fully non-inductive steady-state operation at Q ~ 5 was achieved when additional LHCD
(20-25 MW) was added.
If other H&CD upgrade options equivalent to 20 MW LHCD are considered, it is also
possible to achieve a fully non-inductive Q ~ 5 operation at H98 ~ 1.6.

In the additional analysis performed using METIS, some noteworthy findings are obtained fromseveral comparative studies:With a high plasma confinement enhancement of H98 = 1.6, fully non-inductive Q ~ 5
operation may accessible with the following H&CD upgrade options and at optimized
operation conditions (fGW ~ 1.0, ne0/<ne> ~ 1.3 - 1.5):
o Q ~ 4.9 with NBI/ECRH/LHCD/ICRF = 33/20/20/0 (i.e. Baseline +20 MW LHCD)
o Q ~ 4.7 with NBI/ECRH/LHCD/ICRF = 49.5/20/0/0 (i.e. Baseline +3rd HNB)o Q ~ 5.3 with NBI/ECRH/LHCD/ICRF = 49.5/40/0/0: (i.e. Baseline+3rd HNB + 20 MW ECRH)

341

ITR-18-003



At a modest plasma confinement enhancement of H98 = 1.2, fully non-inductive Q ~ 2
operation would be accessible with the H&CD upgrade options at optimized operation
conditions (fGW ~ 1.0, ne0/<ne> ~ 1.3 - 1.5):
o Q ~ 1.7 with NBI/ECRH/LHCD/ICRF = 33/20/20/0 (i.e. Baseline +20MW LHCD)
o Q ~ 1.6 with NBI/ECRH/LHCD/ICRF = 49.5/20/0/0 (i.e. Baseline +3rd HNB)o Q ~ 1.8 with NBI/ECRH/LHCD/ICRF = 49.5/40/0/0: (i.e. Baseline+3rd HNB + 20 MW ECRH)

Partially inductive (fNI > 0.6) long-pulse operation (~3000 s) with Q = 3-5 may be possible
with several H&CD upgrade options and at H98 = 1.2 - 1.6, if the flat-top plasma current is
optimized for better confinement (i.e. higher Ip than in fully non-inductive operation).

From the plasma stability analysis performed using the KINX and MISHKA stability codes, it canbe concluded that:

Reduction of the internal inductance due to the inherently off-axis LHCD current drive
noticeably reduces the MHD stability limit for n = 1 external modes.

For fully non-inductive plasmas having similar βN and Q ≥ 5, from the plasma stability pointof view, the H&CD upgrade option including both 3rd NBI and a total of 40 MW EC appears
to be equivalent to, and possibly more flexible than, the H&CD upgrade option including
20 MW LHCD.

There are several research areas which require further study and analysis:

Improvement of the physics basis on stable non-inductive operation at high plasma
confinement:o Physics mechanism of the non-linearly coupled core-edge confinement improvement[Garcia, 2015].
o Dynamic evolution of internal transport barriers for their control and/or avoidance.

Stability of non-inductive scenarios against, in particular, NTMs, internal modes and RWMs
together with control requirements for the sustainment of long-pulse operation at high-N.



Optimization of multiple plasma profiles for achieving high plasma confinement (H98 = 1.2 1.6) and improved stability (N ≥ 4li), using the baseline H&CD mix and its possible
upgrades.

As a final summary and recommendation for ITER H&CD upgrade:Both 3rd HNB and 20 MW ECRH power upgrade can provide a non-inductive steady-state
operation capability comparable to the H&CD upgrade option which includes 20 MW of
LHCD.



Further R&D effort will be necessary to establish routine plasma operation at higher
confinement enhancement (H98 = 1.4 - 1.6).

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Appendix F: Physics Analysis and Risks of 5 MA/1.8 T
Scenarios
A detailed analysis of the loads on the ITER TF coils has shown that operation of the TF set at a
toroidal magnetic field of 1.8 T will be possible with plasma currents of up to 5 MA. The main
results and conclusions are summarized in [Mitchell, 2016]. The exploitation of this option becomes
attractive if it is possible to operate the heating and current drive systems at these current and fieldvalues, since it offers the possibility of access to H-mode operation at lower injected powers,
implying that H-modes (and ELM control) might be studied at an early stage of the research
program.
The main results of physics studies of plasma scenarios performed at 5 MA/1.8 T are summarized
in this Appendix. The basic assumptions considered in these analyses are that ECRH heating at 104
or 110 GHz (≤ 6.7 MW) and at 170 GHz (≥ 13.3 MW) will be available and fully absorbed in theplasma, 10 MW of ICRF will also be available and that the magnetic equilibrium during the current
flat-top of these plasmas is essentially identical to that of the 15 MA/5.3 T H-mode scenarios. This
allows, in principle the achievement of both He and H H-modes at 5 MA/1.8 T in PFPO-1 andPFPO-2. Specific studies on heating of these plasmas are being performed including detailed
simulations of ECRH absorption in these modelled plasma scenarios. The initial simulations
performed assumed that 20 MW ECRH heating can be maintained for arbitrarily long periods,
whereas there may be limitations for the low-frequency portion of the ECRH power limiting it topulses lengths of several tens of seconds (see Appendix G).

F.1 Plasma initiation with magnetic field of 1.8 T
Results of the study of plasma initiation with a magnetic field of 1.8 T are described in [Kavin,2016-1] for hydrogen plasmas and in [Kavin, 2016-2] for helium plasmas. The studies have been
performed with the 0-D plasma transport code SCENPLINT [Gribov, 2007], using theelectromagnetic input data obtained with the TRANSMAK code [Belyakov, 2003], assuming Be asa single impurity. In the simulations the Be influx to the plasma was described by physicalsputtering of the Be first wall due to the wall bombardment by hydrogen (or helium) ions and by Beions. After breakdown, an additional gas puffing is required (less than about 1021 s-1, assuming a
total gas volume of 1000 m3) for ramping-up the plasma density after the plasma initiation and forprevention of runaway electron generation.
The studies performed reached the following conclusions:

The narrow range of gas pressure required for ohmic plasma initiation at Bt = 1.8 T indicatesthat its experimental achievement is possible, but very difficult in ITER. The experimental
achievement of ohmic plasma initiation in helium appears to be more difficult than inhydrogen;

ECRH-assisted plasma initiation (at 104 or 110 GHz) broadens the range of helium prefillpressure to 0.2 < p < 0.75 mPa, assuming that 1 MW of ECRH power is absorbed in theplasma. At these high prefill pressures no generation of the runaway electrons is expected;



The value of plasma current at the end of plasma initiation (t = 3.5 s) is approximately
independent of the ECRH pulse duration, ΔtECRH, if ΔtECRH > 1 s. If ΔtECRH < 1 s, the plasma
current at t = 3.5 s decreases with a reduction of ΔtECRH (by about a factor of 2 at ΔtECRH= 0.1 s).

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F.2 5 MA scenarios in a magnetic field of 1.8 T
Design and simulations of 5 MA full bore plasma scenarios at Bt = 1.8 T were performed using the
DINA code [Khayrutdinov, 1993] with feedback and feedforward control of the plasma current,
position and shape, assuming a W divertor and Be first wall, and assuming the separatrix positionand shape during the current flat-top to be similar to those in 15 MA/5.3 T scenarios. It was also
assumed that the maximum value of the current in the central solenoid CS1 coil at the initial
magnetization and at the end of current flat-top is 20 kA (the design limit is 45 kA). The value ofthe initial magnetic flux provided by the PF system with this assumption is ~59 Wb.
F.2.1 5 MA hydrogen ohmic scenarios
The study of the maximum duration of the 5 MA current flat-top in hydrogen ohmic scenarios
limited by 20 kA in the CS1 coil is described in [Kavin, 2016-2]. The blue line in Figure F-1 shows
the duration of the 5 MA current flat-top versus assumptions on the tungsten content, γW = nW/ne,during the flat-top. With increasing γW, the duration decreases from 17 s, at γW = 3×10-5, to 7 s, at
γW = 25×10-5. For comparison the duration of the 3 MA current flat-top is shown by the red line.
Modelled electron and ion temperatures and density profiles in ohmic 5 MA/1.8 T hydrogenplasmas modelled with ASTRA are shown in Figure F-2.
45

T
40

,s

flattopIp =3MA
Ip =5MA

35
30
25
20
15

5

 *10

10

W

5
0

5

10

15

20

25

30

Figure F-1: Duration of the plasma current flat-top in 3 MA and 5 MA hydrogen ohmic scenarios fordifferent assumptions on the W content, γW = nW/ne.

F.2.2 5 MA/1.8T helium H-mode scenario with 20 MW of ECRH
The simulation of a 5 MA helium H-mode scenario is described in [Kavin, 2016-4]. It is assumed
that the plasma is in H-mode, when the power loss to the SOL, PSOL, is higher than the power
threshold value deduced from [Martin, 2008] for deuterium multiplied by C = 1.4 (for helium):
𝑃𝑆𝑂𝐿 > 𝑃𝐻 𝑡𝑜 𝐿 = 0.0488𝐶𝑛𝑒0.72 𝐵𝑡0.8 𝑆 0.94 [𝑀𝑊, 1020 𝑚−3 , 𝑇, 𝑚2 ] ,where PSOL = Pheat – Prad – dWth/dt. The waveforms of Be, γBe = nBe/ne, and W, γW = nW/ne,concentrations assumed in the simulations are shown in Figure F-3.

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Figure F-2: The electron and ion temperatures and density profiles in ohmic 5 MA/1.8 T hydrogenplasmas simulated by ASTRA [Pereverzev, 2002], here at <ne> ~ 0.5nGW. The simulation displays a verystrong electron-ion coupling due to pure ohmic heating, as opposed to the case when 20 MW of EC power
is applied described in section F.2.2 .

0.12

2.5 10

 Be

0.1

2 10

0.08
1.5 10

-5

3

W

Zeff

-5

2.5
-5

2

0.06
1 10

-5

0.04
5 10

0.02
0

1.5

-6

time,s
0

40

80

120

160

200

240

280

time,stime,s
0
0

1
40

80

120

160

200

240

280

0

40

80

120 160

200

240

280

Figure F-3: Assumptions for the contents of Be and W, and the waveform of Zeff obtained in 5 MA helium
H-mode scenario.

An example of the scenario simulation with integrated control of plasma current, shape, plasma
density with 20 MW of additional heating is shown in Figure F-4. During the plasma start-up, the
ECRH (assuming 1 MW absorbed in the plasma) was used only during 2 s for assistance of plasma
initiation, from the gas breakdown (tBD = 1.1 s) until t = 3.1 s. The ECRH power is then switched
off. The X-point is formed at tXPF = 11.8 s (Ip  3.1 MA). At this time the volume averaged densityof He ions is 0.15×1019 m-3, and increases to ~0.4×1019 m-3 at the start of current flat-top(tSOF = 20 s). During 4 s the density of He ions increases to from 0.8 to 1.6×1019 m-3 (t = 24 s), i.e.
the ratio <ne>/nGW increases from 0.2 to about 0.4. At t = 24 s 20 MW of ECRH is applied, leading
to the L- to H-mode transition (at t = 25 s) with a subsequent increase in the density of He ions
during 5 s to ~2.0×1019 m-3 (t = 29 s). This corresponds to <ne>/nGW = 0.5 at the subsequent phaseof the plasma current flat-top. The energy confinement time in H-mode is about 1.7 s. The plasma345

ITR-18-003

density value is kept constant until the end of the plasma current flat-top (tEOF = 265 s) – whichassumes no limitation on the pulse length of the gyrotrons operating at 104 GHz5.
0

2.1

0.5

Ip , MA

-1

a, m

p

0.4

2

-2
0.3

1.9

0.2

1.8

-3
-4
0.1

-5

1.7

time,stime,s
-6
0

40

80

120

160

200

240

280

0
0

40

80

120

160

200

240

280

2.2

2

k

0

40

80

120

160

200

240

280

25

l (3)

2

sep

time,s
1.6

Paux, MW

i

20

1.8

1.8
1.6

1.6

15

1.4
1.4

1

1.2

0

40

2.5

80

120

19

-3

160

200

240

5

0.8

time,s
1

10

1.2

280

0.6

time,stime,s040

80

120 160

200

240

280

5

0

4

0.5

1.5

3

0.4

1

2

0.3

0.5

1

0.2

80

120

160

200

240

120

160

280

5

0

0

1.2

<Ti >, keV

40

80

120 160

200

240

280

240

280

0.1

0

40

80

120

160

200

240

280

240

280

2

n , 10 19, m-3

Pcont /PHL

He

1

4

200

time,stime,stime,s40

80<n e>/n G

e

e

2

0

40

0.6

<T >, keVn , 10 , m

0

0

1.5

0.8

3

1

0.6
2
0.4

0.5

1

0.2

time,s
0

0

40

80

120

160

200

240

time,stime,s
280

0

0

40

80

120 160

200

240

0

280

0

40

80

120

160

200

Figure F-4: Time evolution of Ip, p, minor radius, elongation, li(3), ECRH power absorbed in the plasma inthe 5 MA helium H-mode scenario, volume averaged electron density and temperature, volume averaged
electron density normalized by the Greenwald density, volume averaged ion temperature, volume averaged
helium density, and the ratio of the thermal conductive power loss to the L- to H-transition power thresholdin the 5 MA helium H-mode scenario (assuming PLHHe = PLHD).

5

It is possible that the ECRH power pulse at 104 GHz may be limited to few tens of seconds; this remains subject of
R&D and technical evaluation (see Appendix G).

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The density which provides the minimum power for H-mode access, nLH,min, has been evaluated on
the basis of 1.5-D transport modelling as the density at which the edge ion power flux starts tosaturate with increasing density (i.e. Pi(n)/Ptot ≈ Pi(nLH,min)/Ptot for n > nLH-min). This follows thefindings of ASDEX-Upgrade experiments [Ryter, 2014] that identified the key role that the edge
power loss through the ion channel plays in accessing the H-mode, as described in Appendix B. It isfound that, for 5 MA/1.8 T plasmas, this minimum density is <ne> ~ nLH,min ~ 0.4nGW, as shown in
Figure F-5, in agreement with the global multi-machine scaling evaluations described in Appendix
B, and corresponds to the maximum of the ratio, Pi(n)/PL-H. The actual ratio at which the edge
power flow saturates depends on assumptions regarding the electron and ion transport (ratio χi/χe),but the value of the density at which saturation takes place depends weakly on this assumption,
provided that the overall confinement is assumed to follow the L-mode scaling (this is thenormalizing parameter for the actual values of χi and χe). Therefore, <ne> ~ 0.4nGW can be taken asa good estimate of the density for the minimum H-mode threshold for ITER over a range of main
ion species and fields/currents when q95 = 3, as originally identified for D plasmas in [Ryter, 2014],
and discussed in more detail in Appendix B.

Figure F-5: Fraction of the edge power flow in the ion channel and expected margin to the H-mode
threshold assuming that 50% of the value provided by the Martin scaling has to be conducted by the ionchannel for 5 MA/1.8T He plasmas and 20 MW of ECRH heating.
Figure F-6 shows the electron and ion temperature and density profiles calculated by the METIS

transport solver [Artaud, 2010] and by the 5-D kinetic Qualikiz neural network (NN) for turbulent
transport [Citrin, 2015]. The green region represents the validity region of the Qualikiz NN. METIS
predicts an electron-ion decoupling 𝑇𝑒 /𝑇𝑖 ~2 while the Qualikiz NN predicts 𝑇𝑒 /𝑇𝑖 ~3. This strong
electron-ion decoupling in the plasma central region is due to the fact that these plasmas are heatedby ECRH and have a very low plasma density and thus low electron-ion equipartition.

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Figure F-6: Electron and ion temperature profiles as simulated by METIS (blue) and the Qualikiz Neural
Network (red) for an H-mode helium plasma at 5 MA/1.8 T with 20 MW of ECRH.

Additional studies have been carried out to perform first principle turbulent transport simulations
for these plasmas in order to evaluate energy and particle transport in the core plasma and thepossible consequences for core confinement of these high Te/Ti ratios. This modelling has been
carried out with ASTRA for D plasmas – to overcome the isotopic effect which is not modelledcorrectly by first principle models – using the GLF23 and TGLF transport models, for ECRHheated plasmas with 20 MW of ECRH and <ne>/nGW ~ 0.5. Resulting kinetic profiles are displayedin Figure F-7. These simulations confirm the strong electron-ion decoupling Te/Ti ~2.5-3.0 near theplasma centre (central ECRH is assumed in these simulations). Comparison of these ITER
simulations with results from ASDEX-Upgrade experiments with ECRH-dominant heating shows
comparable scale lengths for density and electron temperature for both GLF23 and TGLF, i.e. R/Lne(~2) and R/LTe (~5), as shown in Figure F-8. For the ion temperature, the scale length evaluated by
GLF23 is much lower than in experiment (R/LTi ~ 2 compared to the experimental value ~ 4-8),
while the TGLF values predicted for ITER are closer to those measured in ASDEX-Upgrade. It is
important to note that in spite of the strong central decoupling between ions and electrons in theplasma centre, density values of <ne>/nGW ~ 0.4-0.5 are sufficient to provide the strong coupling of
the electrons and ions at the plasma edge, and therefore ensure sufficient ion power flow to sustain
the H-mode.

Figure F-7: Density (left) and temperature (right) profiles simulated by ASTRA with GLF23 (dashedlines) and TGLF (solid lines) for D plasmas with 20 MW of ECRH at <ne>/nGW ~ 0.5.

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Figure F-8: Top: scale lengths R/Lne, R/LTe and R/LTi for ITER simulations of ASTRA 5 MA/1.8 T D
plasmas with 20 MW of central ECRH and with anomalous transport modelled with GLF23 and TGLF.
The pink region around r = 1m corresponds to ρtor ~ 0.5. Bottom: ASDEX-Upgrade experimental scale
lengths for plasmas with dominant ECRH heating [Sommer, 2015]. The equivalent scale lengths for
5 MA/1.8 T D plasmas modelled with TGLF are illustrated by pink region on these figures, demonstrating
reasonable agreement.

F.3 Toroidal field ripple for a toroidal magnetic field of 1.8 T
The toroidal field (TF) ripple has been evaluated for the vacuum vessel regular sectors at threelevels of toroidal magnetic field (Bt = 5.3 T, 2.65 T and 1.8 T at 6.2 m) taking into account the
reduced mass (by ~10%) of the ferromagnetic shielding blocks (ferromagnetic inserts) #30-35(without change in their geometrical models). Moreover the measured value of the saturation
magnetization of the steel SS 430 was used in the calculations (μ0Ms = 1.645 T).
The results obtained are reported in [Amoskov, 2016]. Figure F-9 shows the lines of constant valuesof the TF ripple in the vacuum vessel regular sectors at three different values of toroidal magnetic
field. At 5.3 T, the peak value of the TF ripple is +0.338%, at 2.65 T it is –0.538% (the TF ripple isovercompensated by the ferromagnetic inserts) and at 1.8 T it is –1.283% (the ferromagnetic inserts
further overcompensate the TF ripple).
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The effect of TF ripple on the L-H power threshold, the pedestal performance and the ELM
character and frequency has been described in Appendix B, and is expected to amount to a 20-30%
pedestal pressure decrease for 1.8 T plasmas with 1.3% ripple compared to those at lower ripple. A
reduction of the TF ripple can be obtained down to ~ 0.5% by decreasing the plasma radius andshifting the outer plasma separatrix inward by 40 cm while keeping q95 = 3 by increasing the plasma
elongation. This poses no particular problem for He plasmas, which can be sustained by ECRH
alone. However, if such ripple reduction strategy is required to obtain high quality H-modes in
5 MA/1.8 T plasmas, this may have deeper implications for hydrogen plasmas: these require ICRF
heating to access and sustain the H-mode confinement regime, and coupling with such long
distances between the plasma and the antenna (~ 55 cm), which may make ICRF coupling very
challenging and thus H-modes difficult to sustain. Finally, the effects of this level of TF ripple on
RF-accelerated ions and NBI ions need to be quantified by further modelling to determine whether
the associated localized power fluxes on the first wall exceed the power handling limits or not.

Figure F-9: Lines of constant values of the TF ripple in the vacuum vessel regular sectors at various valuesof the toroidal magnetic field at R = 6.2 m: 1.8 T (left), 2.65 T (centre), 5.3 T (right).

F.4 Specific issues on H-mode access and operation at 1.8 T
F.4.1 H-mode physics issues
The issues associated with access to H-mode operation at 1.8 T, and in particular during PFPO-1,
deserve special discussion. The advantage of operation at 1.8 T to access the H-mode in PFPO-1 isillustrated in Table F-1 , which lists a series of possible scenarios with standard assumptions for the
H-mode threshold for H/He/D/DT described in Appendix B. The selected operational density,ne ~ 0.4nGW, corresponds to the expected density for the minimum H-mode power threshold for
ITER. This density corresponds to the multi-machine scaling predictions [Ryter, 2014], and isfound by ASTRA modelling to ensure good sharing of the edge power flow between the electron
and ion channels. Below this density the edge ion power flow decreases with decreasing density. As
described in section F.2.2 and Appendix B, this provides a good guideline for the evaluation of thisminimum density for H-mode access and it is adopted for ITER evaluations.

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Table F-1 – Predictions of H-mode threshold power for several scenarios according to theMartin 2008 scaling and standard ITER assumptions (Density values at <ne> ~ 0.4nGW).
ne,min(1020 m-3)

BT
(T)

S
(m2)

Pth - H2(MW)

Pth - He(MW)

Pth - D2(MW)

Pth - DT
(MW)

0.16

1.8

683

20

10 - 15

10

8

0.25

2.65

683

37

19 - 28

19

15

0.5

5.3

683

106

53 - 80

53

43

Finally, as mentioned in section 2.5.4.9 and Appendix B, the effect of ~10 % helium on hydrogenplasmas H-mode access and performance should be further assessed in order to determine if therecent from recent JET [Hillesheim, 2016] can open a sizeable operational space for predominantly
hydrogen H-mode plasmas in PFPO-1.
F.4.2 ECRH and ICRF heating of He and H plasmas
Exploiting the option for low field operation in ITER requires not only the ability to operate the TF
set at 1.8 T, but also the ability to heat the plasma efficiently. This requires efficient ECRH heating
at 1.8 T. The design frequency of 170 GHz corresponds to 3rd harmonic heating at 1.8 T that, apriori, may not have the highest efficiency; in addition, it does not allow the ECRH to be exploitedfor plasma start-up (no successful use of 3rd harmonic breakdown/ burn-through has been achievedon existing devices). A parallel activity to evaluate the possibility of operation of some fraction of
the ECRH system at both 170 and 104 GHz or only at 110 GHz has, therefore, been launched, as
described in Appendix G. In addition, analysis and design modifications will be required forspecific transmission line (TL) and equatorial launcher (EL) components, as described in Appendix
G. The aim would be to achieve ECRH pulse lengths of several tens of seconds either at 104 GHz
or 110 GHz, which would be suitable for the experimental program foreseen to explore H-mode
access at 1.8 T. It should be noted, however, that a recent study suggests that ohmic breakdown may
be possible at 1.8 T in ITER [Sips, 2017]. Indeed, JET has demonstrated successful breakdown for
Bt = 2.3 T and an electric field of 0.23 Vm-1, confirmed by modelling performed with the DYON
code. This indicates that ohmic breakdown in ITER at 1.8 T may be feasible.
Further ECRH commissioning time will be required if plasma scenarios at 1.8 T are to be exploredin the PFPO-1 campaign. For the 1.8 T scenarios, commissioning of integrated operation of the
ECRH system at 104 or 110 GHz would be undertaken, followed by the use and commissioning of
the two frequencies of the ECRH system by combining 104 or 110 and 170 GHz gyrotrons. Using
ECRH with 104 or 110 GHz frequency (representing a total EC power of up to 6.7 MW, injectingfrom the equatorial launcher) would be essential for 2nd harmonic breakdown, if ECRH start-up
assist were required, as well as for plasma preheating at 1.8 T, if required to ensure adequate
absorption of the main heating ECRH power at 170 GHz. This would provide a maximum absorbed
ECRH power of ~20 MW for the development of the plasma scenario. This scenario will be further
assessed with appropriate dedicated modelling of ECRH power absorption at these two frequencies
integrated into the simulations.
As described in section 2.5.3, the foreseen ICRF scenarios at 1.8 T are 2nd harmonic H heating bothin H and He plasmas. While this heating scheme is known to be efficient in H plasmas being amajority heating scheme, the efficiency of 2nd harmonic H minority heating in He plasmas remains
to be confirmed and further quantified.

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Since increasing the gap between the plasma and the wall is envisaged in order to reduce themagnetic field ripple, comprehensive data on the effect of increased outer separatrix to antenna gap
on coupling of RF power to the core plasma is needed to predict the available absorbed ICRF power
available for two basic plasma shapes proposed for PFPO-1 (standard and displaced inwards by 40cm). The efficiency of ICRF coupling in H plasmas in the 5 MA/1.8 T conditions with reduced
ripple can potentially have significant impact in the split of the operational time between hydrogenand helium H-modes at 1.8T in PFPO-1 and PFPO-2. Experiments during ITER construction needto develop a detailed database of the effect of ELMy and suppressed-ELM H-mode edge plasmas
on coupling of ICRF power. The RF coupling is usually better in L-mode than in H-mode plasmas,
due to the smaller edge density gradients. Optimizing the matching algorithm settings for the L- to
H-mode transition, during which the coupling varies rapidly, will require due attention, since the
H-mode access is only possible with coupled ICRF power for hydrogen plasmas. The acquisition of
sufficient experimental data will allow optimization of the algorithm by better anticipation of thetransition.
A few initial conclusions can be put forward regarding the ICRF coupling efficiency for these ITER
5 MA/1.8 T plasmas from the observations in JET experiments: 1.8 T H-mode scenarios have avery low density and a low collisionality, as discussed in Appendix B and section 2.5.4.9. Hence,
the high energy tail of RF-heated ions may create high local power loads on the first wall, even
without considering magnetic field ripple. This behaviour has been observed in JET for low density
plasmas, as illustrated in Figure F-10 [Lerche, 2015-1].

Figure F-10: Increase in the neutral particle analyzer signal for a range of ICRF powers versus plasma
density. This increase is due to poor confinement of RF-heated ions at low fields, currents and densities.

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Figure F-11: ICRF coupling resistance as a function of the Gas Injection Module (GIM) - strap distancefor mid-plane and top gas injection in JET.

ICRF coupling properties are very dependent on the value of the plasma density and on the locationof the gas injection modules. In ITER, gas injection at the top is foreseen to fuel the plasma and to
improve ICRF coupling performance. As described in section F.3, increasing the plasma-wall
distance is the option to reduce the separatrix value of the magnetic field ripple at 1.8 T, and thismay degrade the ICRF coupling efficiency. JET experiments have shown that the coupling does not
significantly depend on the GIM-strap distance when using top injection6 [Lerche, 2015-2], as
illustrated in Figure F-11, which displays the coupling resistance versus the GIM-strap distance for
top and mid-plane gas injection at JET. These observations suggest that the use of top gas fuellingin ITER should be explored in order to maintain good ICRF coupling if the plasma-antenna distancehas to be increased to reduce TF ripple in 5 MA/1.8 T H-mode plasmas. Using a local midplane gas
injection system instead of top injection is more likely to increase the coupling efficiency
significantly in these conditions, as illustrated in Figure F-11. This modification of the gas fuellingcapabilities is under assessment within the project [Schneider, 2017-2].
From the ICRF commissioning point of view, 1.8 T scenarios would be advantageous to test RF
capabilities earlier in the research plan, particularly in H H-mode plasmas.

F.5 Physics risks
There are a number of physics risks to the success of these H-mode scenarios, associated with
operation at 1.8 T, that need to be carefully considered and evaluated: Plasma heating by ECRH: for the 170 GHz ITER gyrotrons at 1.8 T, 3rd harmonic, or X3, central
heating is available and has been demonstrated to work well in present-day devices, both in TCV
plasmas preheated by X2 heating [Alberti, 2005] and in ASDEX Upgrade plasmas without X2
preheating [Hohnle, 2011]. Evaluating the necessity (or otherwise) of X2 preheating in ITER
will require further simulation studies, but dual-frequency gyrotron operation if implemented forassisted ECRH breakdown/burn-through would, nevertheless, allow such a scenario in ITER.

6

The worst case for ICRF coupling performance is divertor injection, the medium case is top injection and the best caseis local injection from, e.g., the equatorial plane.

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 EC start-up assist: reliable plasma start-up in ITER may require an assist from up to ~ 5 MW of
ECRH, particularly at 1.8 T, where the connection lengths in the proximity of the toroidal nullare shorter, unless ohmic breakdown can be achieved. Since the use of ECRH at 170 GHz in theX3 mode is not feasible due to low absorption at low temperature and density, start-up assist
would require X2 ECRH heating at 104 or 110 GHz.
 Toroidal field ripple: the peak toroidal field ripple on the separatrix at 1.8 T for regular sectors(i.e. not NBI vacuum vessel sectors) is estimated to be -1.28%, as compared to -0.54% and0.34% for 2.65 and 5.3 T respectively, as described in section F.3. At this level of ripple, reduced
H-mode performance has been observed in present days tokamaks for low pedestal collisionality[Saibene, 2007; Saibene, 2008], although a much smaller effect is found on PLH [Urano, 2006;
Andrew, 2008]; further discussions on this topic can be found in Appendix B. If the level of
1.3% ripple at 1.8 T were found to degrade H-mode performance to an unacceptable level in
ITER, one option for decreasing the level of ripple is to decrease the minor radius and displace
the plasma radially inwards. To achieve a ripple at the separatrix of ~0.55% requires an inward
displacement of the outboard mid-plane separatrix location from 8.2 to 7.8 m [Kavin, 2017] andadjustment of the plasma shape to keep q95 ~ 3. Although this plasma scenario has been shown tobe controllable from the plasma position and vertical stability points of view [Kavin, 2017] thelarger distance of the plasma to the ICRF antenna and to the ELM control coils poses specificchallenges. Increasing the distance of the plasma separatrix to the wall by ~0.4 m (i.e. from ~0.16 m to 0.56 m) may strongly decrease the coupling resistance and this is likely to reduce thecoupled ICRF power to very low values for such plasmas. The increase of distance from theplasma to the wall also significantly increases the distance from the ELM control coils to theplasma, which is typically 0.5-0.9 m for the un-shifted plasma position, and this reduces themagnitude of the field from the coils at the separatrix by a factor of ~ 1.5-2.0 depending on thecoil location (Upper, Lower or Equatorial). Thus, the current to be applied to obtain a similarvalue of radial magnetic perturbation at the plasma surface to that in the reference separatrix
position needs to be increased substantially.
 Low density operation: ITER’s reference operating point for Q = 10 is at a plasma current of
15 MA and a line averaged density of ~ 1020 m-3. However, the empirical H-mode density limit
provided by the Greenwald density depends linearly on plasma current, and thus the equivalent
operating point at 5 MA corresponds to ~ 0.351020 m-3 (i.e. at 85% of the Greenwald density).
As discussed in Appendix B, experimental results indicate that there is a density minimum inthe power threshold scaling for a wide range (but not all) of plasma conditions. A possible
explanation for this density minimum is the thermal decoupling of ions and electrons at theplasma edge [Ryter, 2014], leading to a density for minimum power to access H-mode in ITERof ne ~ 0.4nGW, corresponding to a density of ~1.61019 m-3 at 5 MA/1.8 T, in these ITER
conditions. In addition, 5 MA/1.8 T operation at plasma densities of ne ~ 2.01019 m-3 in
PFPO-1 and PFPO-2 creates two issues for H-mode operation:– The poor collisional coupling between ions and electrons and the dominant electron heatingin ITER leads to very large high radial Te gradients in the plasma core and significant
decoupling of the electron and ion channel, which may affect the energy confinement of the
H-mode plasmas in this phase;– The low density operation in hydrogen plasmas leads to very low absorption of the heating
neutral beam (HNB) and thus to large shine-through loads on the first wall and shield block
panels. This restricts the power that can be injected in this type of plasmas with NBI in
PFPO-2 or the length of the NBI heating pulse, particularly for hydrogen plasmas. On the354

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basis of recent assessments [Singh, 2017; Kim, 2017] the maximum power for long-pulse
HNB operation in hydrogen plasmas with ne ~ 2.01019 m-3 is 8 MW while for helium
plasmas this is 16.5 MW, where the energy of the NBI ions is reduced to 500 and 650 keV
respectively in order to decrease the shine-through loads.
 Coupling of ICRF power with high ripple levels and through the L-H transition for hydrogenplasmas: Using ICRF with a minority heating scheme leads to the formation of high energy ion
tails and ions that can be trapped in the relatively high ripple (~1.28%) wells. Although thescheme foreseen for ICRF heating of hydrogen plasmas is not one which leads to very highenergy tails (second-harmonic majority heating) the expected loads due to ICRF heating in thesehigh ripple conditions need to be evaluated. Another important issue is that ICRF heating isessential to access the H-mode regime for hydrogen plasmas and, thus, good ICRF couplingneeds to be maintained throughout the L-mode, L-H transition and H-mode phases. While
resolving the issues of ICRF coupling to plasmas in H-modes with controlled and suppressed
ELMs is required for the application of ICRF to ITER high-Q scenarios, the specific issue ofmaintaining good coupling through the L-H transition is critical for hydrogen H-mode operationin PFPO-1. In PFPO-2, the situation is somewhat less critical due to the possibility of using NBI
heating at reduced power during this phase. Previous studies have shown that it is possible tomaintain sufficient plasma coupling through the L-H transition to provide ~20 MW of ICRF
heating in the access phase to Q = 10 plasmas; a similar assessment will be repeated for the Hmode access phase of hydrogen H-modes at 5MA/1.8T in the near future.
 ELM control and MHD control: Some of the baseline systems for MHD control will not be fully
available in PFPO-1. For instance the set of power supplies for the ELM control coils will only
allow a restricted set of 3-D field configurations to be produced and at a current level in thecoils which will be much lower than in the baseline configuration, where every coil will beconnected to its own power supply (e.g. fixed helicity n = 3 with maximum current level of~30 kAt is foreseen for PFPO-1). This will require adjustment of the alignment between theapplied magnetic perturbation and the plasma edge magnetic field through the adjustment of q95
in a similar way to current tokamak experiments. This substantially differs from the operationalmode planned for PFPO-2 and FPO, where all power supplies will be available and the appliededge magnetic field perturbation will be adjusted to the value of q95. The βN of the H-mode
scenarios is expected to be low (~1.2) and therefore it is not clear if the performance in theseplasmas will be strongly affected by neoclassical tearing modes. If this is the case, it should be
noted that the flexibility for NTM control in PFPO-1 will be reduced compared to that during
PFPO-2 and FPO because of the reduced set of heating schemes in this phase. If a significant
amount of off-axis ECRH/ECCD is required to control NTMs in these plasmas, this will reduce
the available central heating and will most likely affect the performance of these H-mode
plasmas, possibly leading to problems of W accumulation, if the remaining level of central
heating is not large enough.

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Appendix G: Heating and Current Drive Staging
The Heating and Current Drive systems in ITER are comprised of the Neutral Beam system forheating & current drive (HNB) and diagnostics (DNB), Ion Cyclotron for heating & current drive(ICH&CD) and the Electron Cyclotron for heating & current drive (ECH&CD). The combination ofthese 3 systems is designed to provide 73 MW of power for the PFPO (H/He) phase of operation,and up to 110 MW of injected power in the fusion power operation phase when including available
upgrade options, which potentially range from a choice of a combination of an extra 20 MW of
ECRH, 20 MW of ICRF, 16 MW of NBI or up to 40 MW of Lower Hybrid power (LHCD). The
total installed power after upgrades can be up to 130 MW.
The current ITER Research Plan is based on PCR-738 that establishes the ITER Baseline[ITER_D_TL5BR3, 2016] and foresees the availability of 6.7 MW of ECRH for the First Plasma(FP) followed by the completion of the installation and operation of the ECRH system up to 20 MW
for PFPO-1. In the current Baseline both the NBI and ICRF systems will be functional for the
PFPO-2 experimental campaigns. The ITER Research Plan is elaborated on the basis that a further
10 MW of ICRF heating may be available within PFPO-1. Feasibility studies into accelerating the
ICRF system to provide 10 MW for PFPO-1 are underway and are reported upon below.
This Appendix covers the following issues for the heating and current drive systems, which are key
for the ITER Research Plan within the staged operation approach:
1. The proposal to operate at 1.8 T and the impact, primarily on the ECRH system;
2. The acceleration of 10 MW of ICRF power, to be made available for the PFPO-1 campaign;
3. The status of the investigation into acceleration of the NBI systems;
4. The upgrade possibilities of the heating systems, including the status of the LHCD system.

G.1 Implications of 1.8 T plasma operation for H&CD systems
Within the elaboration of the ITER Research Plan, the participant experts recommended toinvestigate the possibility of considering 1.8 T operation as this may enable access to H-mode
operation within PFPO-1 and offer an H-mode scenario (5 MA/1.8 T) for which uncontrolled ELMs
are not predicted to cause divertor monoblock melting in this and later phases of the research
program (see Appendix C and sections 2.5.4.9 and 2.5.5.7). At its 21st meeting in October 2016, the
STAC recommended further study of the implications of 1.8 T plasmas, both in terms of plasma
operation and in relation to the performance of ITER auxiliary systems (especially for heating andcurrent drive and diagnostic systems).
Operation at 1.8 T has primarily an impact on the ECRH system. The ECRH system is available forboth PFPO-1 and PFPO-2 plasmas, and 1.8 T operation can have impacts on hardware, design, andrequires additional physics analysis in order to maximize the efficiency of operation at this field.
Regarding the possibility of combined heating for PFPO-1, taking advantage of the accelerated
program for the ICRF system discussed below, it is not envisaged that any hardware changes will
be needed. Therefore, there is no impact on the ICRF system hardware and the studies are centredon the optimization of the use of the baseline ICRF hardware for heating of 1.8 T plasmas, as
discussed in Appendix F. Similarly the possible operation of 1.8 T with NBI in PFPO-2 does nothave hardware implications for the NBI system; the main issue is here is to develop plasma
scenarios with acceptable shine-through loads on the blanket shield block modules. This may
impose limitations to the length of the NBI heating pulse and/or to the maximum ion beam energy356

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that can be used during this and, possibly, later phases (see 2.5.5.7 and 2.6.3.6), none of which
impact the NBI hardware. The hardware of the ICRF and NBI systems is, therefore, not affected by
1.8 T operation and these systems are not discussed further in this section.
The ECRH system is designed for optimized operation at the nominal field of 5.3 T, and to provide
a similar functionality at half field (2.65 T), in addition to reduced functionality at intermediate
fields. The present design is compliant with the requirement for depositing the 20 MW acrossnearly the entire plasma cross section. It includes the potential for a power increase of up to
40 MW, with the addition of 24 1 MW gyrotrons, associated power supplies and transmission lines,
located either in an extension to the RF Building or in the Assembly Hall. Note that no changes are
required within the Tokamak Building for an ECRH upgrade option, as the additional gyrotrons andtransmission lines (TL) will be connected to 56 entries available between the four upper launcher
and one equatorial launcher.
This section describes the additional requirements imposed on the ECRH system for operation at
1.8 T, and the subsequent impact on the ECRH system design.
G.1.1 Impact of 1.8 T plasma operations on the ECRH system
Including 1.8 T plasma operation in the IRP (one-third nominal field) would require the ECRH
system to provide the following functions:

Up to 6.7 MW for Plasma start up using either O1 or X2, if required for plasma breakdown
and burn-through;

20 MW of central heating in a combination of either O1, X2 and/or X3;

consistency of the inclusion of the above requirements without sacrificing the functional
requirements in the range of 2.65 T operation during the PFPO-1, PFPO-2 and FPO
operating phases.

The primary impact on the ECRH system is driven by the need to provide breakdown and burnthrough at one-third nominal field should this be required, as preliminarily evaluated by modellingin Appendix F. In order to meet this requirement, an additional lower frequency gyrotron operation
mode is needed to provide these functions via either the O1 (~60 GHz) or X2 (~110 GHz) using theexisting transmission lines (TL) and equatorial launcher (EL). Note that the upper launcher (UL)
has a narrower frequency bandwidth not compatible with the above frequencies. The TL and EL
offer broadband operation, but the range of 60 GHz is far outside the bandwidth, therefore only theX2 path is being investigated.
In order to maintain functionality at the 2.65 T operation (O1) a dual-frequency gyrotron isassumed, so that a set of up to 8 gyrotrons could either provide the X2 heating (~110 GHz) or the
O1 heating (170 GHz). The Toshiba 170 GHz gyrotrons have already demonstrated operation atfour frequencies (104, 137, 170 and 203 GHz), and the Japan-Domestic Agency (JA-DA) gyrotrons
are configured for the EL steering mirrors that provide the breakdown and burn-through functions.
Therefore, a design proposal utilizing the Toshiba tubes at either 104 GHz or 137 GHz, with a backup option of using the 110 GHz tubes used for JT60-SA, which are compliant with thesuperconducting magnets and matching optics units of the 170 GHz gyrotrons, is being developed.
Preliminary ECCD analysis was performed by L. Figini and D. Farina (IFP-CNR) assessing thedeposition ranges at the two frequencies as a function of the toroidal field and comparing these withthe 170 GHz functionality, as shown in Figure G-1.

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Figure G-1: (left) Comparison of the radial deposition location of the EC beam as a function of the toroidal
magnetic field for the 170 GHz and 104 GHz frequencies and (right) the 170 GHz and 137 GHz frequencies.
The deposition range is calculated for the end of flat-top, and the density and temperature profiles scaled
proportionally with the magnetic field relative to the H-mode 15 MA / 5.3 T scenario.

The 104 GHz option has similar deposition ranges for the 1.8 T and 2.65 T operating ranges, whilethe 137 GHz option corresponds to poor central deposition for both the 1.8 T and 2.65 T ranges.
Therefore, a path pursuing a dual-frequency 104/170 GHz operation is proposed, with the back-up
of dedicated 110 GHz gyrotrons that would be replaced in the active phase once the 1.8 T operation
has been concluded.
Note that the EC system has been designed for a single-frequency system, as the 170 GHz
frequency achieved the required functionality at 2.65 T and 5.3 T considered in previous versions of
the ITER Research Plan. Introducing the dual-frequency operation has a significant impact on the
gyrotrons, TL and EL, which will be outlined in the following sections. Note that the upperlauncher (UL) has a four mirror quasi-optical system that limits the operating bandwidth to anarrow region around 170 GHz, with the 104 GHz far outside this operating window.
G.1.1.1 Impact on Gyrotrons
The primary impact on the gyrotrons is the development of a 104 GHz/170 GHz dual-frequency
gyrotron (DFG) with output power ≥0.8 MW and pulse lengths in the range of several tens ofseconds. The present 170 GHz gyrotron procurement remains unchanged with the delivery of theeight tubes maintaining the present schedule. In parallel, the DFG would be developed using thepresent 170 GHz design, but with a three phase optimization, as illustrated in Figure G-2. There is ahold point at each phase that assesses the output power, mode purity and generated stray power. If

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the results indicate that the DFG parameters mentioned above cannot be achieved, then the
mitigation path of procuring 110 GHz tubes (JT60-SA) will be implemented.

/ QST

/ QST

Figure G-2: Description of the different phases for the dual-frequency gyrotron study.

The first two phases will cover the new design of internal critical components: internal launcher andquasi-optical mirrors (see Figure G-3), and the last phase is focussed on the external matching optics
unit and design finalization.
The most critical technical issue for developing a dual-frequency tube is to obtain a high efficientconversion from the excited mode in the cavity (either TE19,7 for 104 GHz or TE31,11 for 170 GHz)to the free space Gaussian mode (TEM00 for both frequencies) via the internal launcher. Then the 4internal mirrors reshape and direct the beam out the diamond window with both frequencies having
a common co-linear path; such that the two additional external mirrors can couple to the HE 11
waveguide (not shown on the picture).
This study aims at finalizing the design of a dual-frequency tube with the following characteristics:
1. Transmission efficiency at diamond window: > 98% (both 104/170 GHz);
2. Gaussian mode at the diamond window: > 95% (both 104/170 GHz);
3. Coupling efficiency of more than 95 % from the TEM00 into an HE11 mode at the entranceof the waveguide (both 104/170 GHz);
4. Power transmission efficiency at the matching optics unit (MOU) of more than 96 % (both104/170 GHz).

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Figure G-3: Sketch of main parts inside the gyrotron showing the critical components to be modified for adual-frequency tube.

Figure G-4: Model of a gyrotron installed in its superconducting magnet (SCM) and connected to
the waveguide through the MOU.

The expected performances will be given at the end of the study, but should not be less than
800 kW at the output of the MOU:

for a pulse length of 100 s at the frequency of 170 GHz;

for a pulse length of 30 s at the frequency of 104 GHz.

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The development program is scheduled to last 18 months, starting from the signature of a task
agreement between ITER Organisation and JA-DA (QST). For reference, a schematic of a gyrotron
installed in its superconducting magnet (SCM) and connected to the waveguide through the MOU isshown in Figure G-4.
G.1.1.2 Impact on Transmission Lines
The transmission line (TL) diameter and its corrugations have been optimized for the frequency of
170 GHz. The operation at a lower frequency (be it 104 or 110 GHz) would result in increased
power densities primarily near the mitre bends. The lower frequency will generate higher losses at
the mitre bends than those for 170 GHz and these losses will be absorbed in the shorter length of theneighbouring waveguide. The resulting power density is typically a factor of 4 higher compared to
the 170 GHz case. The design of these components will not be modified at this stage, but anassessment will be performed based on their finite element models to determine the limitingtransmitted power and/or pulse length to determine the maximum power handing and pulse length
capabilities.
This assessment will require additional resources to re-analyze the finite element models of eachcomponent and to assess the limiting power/pulse length possible for the TL. Note that no increaseof total loads is envisioned to the Component Cooling Water System that serves the ECRH system(CCWS-2A or CCWS-1).
G.1.1.3 Impact on Equatorial Launcher
The operation of the 104 GHz (or 110 GHz) system will be limited to the equatorial launcher as
mentioned above, as the upper launcher quasi-optical design has a much narrower frequency
bandwidth that is only a few tenths of GHz in width. The impact on the equatorial launcher includes
the following:
1) Diamond window: The window has a thickness of 1.11 mm, consistent with the 170 GHz
operation, sufficiently thick to withstand all over-pressure load conditions and kept to aminimum to avoid cost increases (proportional to the volume of the disk) and minimize thewindow absorption coefficient. In order to transmit at both 104 GHz and 170 GHz, the disk
thickness has to increase to ~1.5 mm (corresponding to integer half wavelengths for both
frequencies). A development program for the qualification of the thicker disk andsubsequent procurement of eight windows would be required;
2) Equatorial Launcher: The compatibility of the existing optical configuration has to beassessed for 104 GHz transmission. The concern is that the beams expand more rapidly,
potentially being too large at each of the two free space mirrors. In this case, the opticaldesign would have to be redesigned for the two frequencies.
Note that the above changes would be implemented for the PFPO-1 operation and be removed after
PFPO-2. The upper launchers cannot be used at all at the frequency of 104 (or even 110) GHz as
the stray field would be too high and would damage components.
G.1.1.4 Impact on Resources
The additional complexity to the EC system and corresponding management of the designmodifications and implementation requires additional engineering resources in the corresponding
section of the IO within the Heating systems Division during the period of 2018 to 2037. This361

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covers the period in which the design, installation, operation and removal of the components for the
DFG operation takes place.

G.2 Acceleration program of 10 MW of ICRF power to be made available for the
PFPO-1 campaign
The Record of Decisions of the 19th meeting of the ITER Council’s (IC-19) states:
7.2 The ITER Council:
i. […]
ii. Requested that the IO, Members and their Domestic Agencies (DAs) accelerate the Neutral
Beam Injection (NBI) and Ion Cyclotron Resonance Heating (ICRH) systems, making use ofexisting expertise in the Members and their DAs and of progress that will be achieved in the
coming 2-3 years, with a view to enabling the use of the ICRH system for the Pre-Fusion PowerOperation-1 in the interests of shortening the period from First Plasma to DT Operation;
The IO and respective DAs are investigating the feasibility of accelerating the ICRF system with aview to providing 10 MW of ICRF power in the PFPO-1 experimental campaign, as has been
assumed for the development of the ITER Research Plan. The report here outlines the current
Baseline and plan of work to determine the possibility of accelerating, in particular, the ICRF
antenna, which is on the critical path for the ICRF system.
The ICRF system is divided into the following procurements:
1.
2.
3.
4.
5.

RF Sources and HV PS;
HV PS;
Transmission lines and matching units;
Antenna;
Control system.

The current high-level planning of the ICRF system based on the 4-stage approach in PCR-738[ITER_D_TL5BR3, 2016] is shown below in Figure G-5.

Figure G-5: ICRF system Baseline schedule based on PCR-738.

The staged operation approach has been developed with the inputs of the Configuration Workshop
and the cost constraints of the DAs for the relevant procurements. Issues have been identified withthe delivery of some of the transmission line components, as current cost constraints and schedule362

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baselined under PCR-738 are not compliant with the accelerated program. However, the main
critical path to ensure 10 MW for PFPO-1 is the antenna, which is currently scheduled to be readyin 2030. This section deals with the antenna acceleration only, as detailed discussions have only
started on the TL deliveries. The main assumptions used as a basis of this evaluation are listed here:The present procurement strategy is adhered to: need to have technical knowledge of design,
knowledge of industry and market for the manufacture and procurement aspects of theantenna project (Note: risk reduction). Therefore, most of the work on design and prototypes
is planned to be carried out by the responsible DA (via task agreements);



Corresponding procurement rules have to be followed;



No new budget or resource is assumed before second quarter of 2018;Assume suitable resource and expertise is available as and when needed;

No Faraday Screen (FS) mock-ups will need to be manufactured before the prototype. As
explained in the next paragraph, part of the proposed acceleration is based on accelerating
the critical path work (RF windows). The FS work then becomes the critical path.

The progress made in order to evaluate the feasibility of antenna acceleration is under evaluation
with the intention of providing a final report to the IC-22 meeting in June 2018.
In order to accelerate 10 MW of ICRF power (one antenna), so that it is available for Assembly
Phase II, would imply that the ICRF antenna would need to be delivered to the IO site by October
2027, thus allowing for six months of testing in the Port Plug Test Facility (PPTF) and three months
for the installation, i.e. this would mean that the antenna would need to be installed and ready for
the Integrated Commissioning II activity in July 2028. Therefore a three-year acceleration relative
to the current antenna schedule (current delivery foreseen in September 2030) would be required.
The current planning of the antenna is shown in Figure G-5 above. There are two main phases, thedesign phase and the procurement/manufacturing phase.
G.2.1 Design Phase
In evaluating the acceleration of the ICRF antenna in the design phase, the critical path activity forthe design and development of a suitable window was considered. The window development could
potentially be part of an accelerated program, by combining the contracts that are currently foreseen
for demonstration of manufacturability and for production of a prototype to demonstrate
functionality. If combined, this would incur a significantly higher cost since the risk to the supplierwould be considered higher than the current two-step strategy.
Removing the window from the critical path of the design phase would then result in the Faraday
screen being the new critical path. The Faraday Screen mock-ups reported at STAC in May 2014
were planned to be repeated, based on the change of material from copper to stainless steel. Fromthe plan it is clear that if the previous mock-up tests for the rectangular cooling could be used tovalidate the new cooling channel proposal, bearing in mind that a prototype will in any case bemade, then the removal of this activity combined with the window acceleration will save significant
time.
The above saving combined with a final acceleration step in the design phase to partially parallel
the Final Design Review (FDR) with the Procurement Arrangement preparation (PA) could save
~ 1 year.

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G.2.2 Manufacturing and Procurement Phase
Acceleration for the manufacture and procurement phase is currently difficult to evaluate as thedetailed procurement strategy has not yet been defined. However, it is assumed that there may bemargin for limited acceleration by optimizing the contract strategy. Currently, without further
detailed investigation, a six-month acceleration is assumed possible, although an increase inresources will be necessary in such a strategy.
While the above acceleration possibilities may be feasible, it should be remembered that there are
currently a number of open issues in the antenna design which need to be addressed, some of which
are highlighted here:
There is a suitable process, which can be qualified, for the joining of titanium to StainlessSteel (TI/SS): The current ICRF antenna design has a requirement for the joining of titanium
to stainless steel but a straight joint between these two materials is not possible due to thecreation of brittle intermetallics. There is some literature available, which suggests that theuse of a metallic interlayer for this type of joint is possible with non-trivial mechanicalproperties. Metallic interlayers to join the titanium to stainless steel are, therefore, being
investigated in two processes namely: Ti/SS friction welding and uniaxial diffusion bonding.
Initial results for the friction welding look promising and, therefore, if the feasibility of thesewelding trials proves to be successful, a welding procedure will be produced. If, instead, atechnique for joining titanium to stainless steel is not found, then the RF antenna design will
have to be altered;

The copper plating process is suitable and can be qualified: The ICRF antenna has someareas within its design which have to be copper coated. However, there is currently nosuitable process which can be qualified. The R&D work on the baseline technique, whichused copper electroplating for the copper coating of Titanium Grade 2, provided
questionable results in both quality and repeatability. There was severe degradation of
coatings including Copper lift-off in the presence of braze material (Ticusil/Incusil). An
alternative technique, proposed after consultation with appropriate experts, suggested cold
spraying techniques as a possible alternative. An R&D contract is underway for cold
spraying, the current results look promising;The titanium embrittlement is demonstrated not to be an issue: The antenna design includes
Titanium which is not a commonly accepted material for ITER vacuum applications. It isparticularly susceptible to H-embrittlement phenomena, brought about by the ITER waterchemistry in the nuclear phase. An evaluation of H-embrittlement effects on the Removable
Vacuum Transmission Line (RVTL) materials, in particular the Ti and the Ti-SS joints, isneeded to gain the required vacuum compliance (VQC1A) approval and a contract has beenplaced to start this work;

The manufacture of the entire window assembly is possible: The window design,
incorporated in the RVTL of the ICRF antenna, is very complex. The design needs to bevalidated by building and testing a suitable prototype window. If issues are found, there is arisk that an alternative window solution will have to be found;

There is an RF test facility available for the tests when required;

The current design activity will be successful in producing a fully compliant design.

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G.2.3 Additional Risks of an Accelerated Programme
Additionally, other risks need to be considered for an accelerated program and these include:

As indicated above, there is still a question on the viability of the window design which
presents a risk to the overall project timescale. The consequences of this risk are larger for
any proposed accelerated program;An accelerated window program is likely to imply higher risk and, therefore, higher cost;

Risk of insufficient resources being available to perform parallel work on the Procurement
Arrangement preparation and Final Design Review;

Risk of insufficient resources being available to prepare and follow up all procurement
contracts. The exact amount of additional resources needed is still to be evaluated.

The possibilities examined to date would result in an eventual acceleration of approximately
1.5 years. If possible to implement, the delivery of the antenna would be then occur at the end of2029, i.e. during the PFPO-1 experimental campaign, and its installation would require a special
shutdown to allow some commissioning and operation during PFPO-1. Such acceleration alsocomes with associated risk and increased human resources. No cost assessment has yet been carriedout in relation to risks and resources, and the possibility of achieving the required acceleration toallow 10 MW ICRF to be installed before the start of PFPO-1 remains the subject of analysis. In
this respect, it should be noted that further discussions with DAs having the relevant expertise inthis area are ongoing to support the analysis of the proposed acceleration activity.
As mentioned previously, a final report will then be made available at the IC-22 meeting in June2018 and will include the full risk assessment and required additional budget and resources for thisacceleration.

G.3 Status of the investigations into the acceleration of the NBI systems
As for the ICRF system, IC-19 has requested a study of the feasibility of accelerating the NBI
systems. The current schedule for the HNB systems is shown below in Figure G-6. Note that theschedule for the DNB is the same.

Figure G-6: Baseline schedule for the Neutral Beam Test Facility (NBTF) and NBI system with respect tothe ITER Baseline.

The success of the NBI systems for ITER relies heavily on the development program that isunderway at the Neutral Beam Test Facility (NBTF), which was considered necessary in order to365

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mitigate the design, manufacturing and operational risks of the injectors. With the current planning,
the timelines in shown Table G-1 are envisaged.
Table G-1 – Experience to be gained in the NBTF facilities prior to ITER NBI operation
Time Line
SPIDER start of experiments 2018

Experience
Four years’ experience on the ion source prior to
1 MV beam experiments

MITICA start of experiments 2021
HNB/DNB to inject into PFPO-2

Ten years’ experience on a prototype injector
prior to ITER operation of the injectors

In order to have confidence in the above schedule it has been decided to investigate the technicallyshortest achievable schedule of the NBTF. With the involvement of all of the stakeholders, the
NBTF schedule is being developed, and this is based on a deliverable, resource loaded schedule, toprovide the input parameters for the beam program at ITER in time for manufacture andprocurement of the critical elements of the HNBs.
A roadmap has been developed, which incorporates the risks that could be encountered and thepossible modifications from one facility to the next, to result in the final design of the HNBs and
DNB. This roadmap forms the basis of the technically fastest achievable schedule from the test
facilities of the NBTF through to operation of the beams on ITER. The development of thisroadmap has taken into account the expertise both of the DAs who contribute to the NB for ITER
and also that of experts in other DAs. It has been concluded that, while some collaborations in theoperation and data analysis are possible and welcomed, no significant acceleration can be achievedin the technically shortest schedule which meets the Staged Approach. The work will be
continuously reviewed in order to determine if there are any areas which could potentially assist
with an accelerated schedule.

G.4 Upgrade possibilities of the ITER H&CD systems (including the status of the
LH system)
The Project Requirements [ITER_D_27ZRW8, 2014] outline the possible heating and current drive
upgrades in section 5.4.3. These upgrades ensure the flexibility of the machine for the different
operating scenarios and cover the possible upgrade of all three (construction) baseline H&CD
systems, i.e. ICRF, ECRH and NBI, but also define the possible upgrade to incorporate a Lower
Hybrid system as a potential upgrade path to increase the capabilities for off-axis current drive.
These possible upgrades are considered in the following sections.
G.4.1 Upgrade of the ECRH system
The present ECRH system shall provide a total of 20 MW of RF power at the frequency of
170 GHz in the plasma. The EC system is composed of High Voltage Power Supplies (HVPS), RF
power sources (or gyrotrons), the transmission lines (TL), the equatorial (EL) and upper (UL)launcher port plugs plus auxiliary sub-systems and services, control systems and test facilities. The
layout is illustrated in Figure G-7, which shows the three buildings that house the ECRH system: RF
Building (PS, gyrotrons and TL), Assembly Hall (TL) and Tokamak Building (TL and launchers).

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The ECRH system has been designed to be compatible with increased power, which can beachieved in two options. First, the power supplies have been dimensioned to be compatible with afuture upgrade of the gyrotron tubes in the range of 1.2 to 1.4 MW (depending on the gyrotrons’required current/voltage requirements and overall efficiency). This can provide an additional
4 - 8 MW with very little changes in the existing infrastructure, as all ECRH components arecompatible with this rating and the cooling pipes are dimensioned for the corresponding increased
flow.

Figure G-7: Layout of the ECH&CD system.

The second option assumes duplicating the ECRH system outside the Tokamak Building thusincreasing the total available power for plasma heating by ECRH to 40 MW. This requires the
additional space, 12 power supply sets, 24 gyrotrons, 24 TLs leading up to the Tokamak Building
and the associated ancillaries. The TL and launchers within the Tokamak Building do not change,as there are 56 entries (24 in the EL and 32 in the UL). The routing and switching mechanisms in
the Assembly Hall will be modified to best distribute the total 48 output beams from the gyrotrons
to the 56 entries based on the HCD physics requirements.
Note that a combination of both options could be implemented to bring up the injected power
between 48 to 56 MW. Obviously, this is strongly dependent on the continued improvements ingyrotron technology, which should not be taken for granted.
G.4.1.1 Location of the 20 MW upgrade of the ECRH system
A 20 MW upgrade would require equivalent space to the one for the existing plant in the RF
Building (B15). PCR-496 [ITER_D_ETLLQC, 2013] identified two options for providing thisspace, as illustrated in Figure G-8:
1) An extension to the RF Building (B15), as represented in the blue shaded area;
2) To install it in the Assembly Hall with an area equivalent to the RF Building, but placed
sufficiently far from the tokamak to avoid the stray magnetic field to impact the gyrotrons.
The most cost effective solution is to utilize the space in the Assembly Hall to avoid the additional
cost in construction of a building extension. The following assumes that the Assembly Hall option
is adopted for the 20 MW upgrade.
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A metallic structure is required that mimics the three levels of B15 and would fit between the south
side of the building, the metrology lab (to the east) and the Collective Thomson Scattering (CTS)
gyrotrons to the north-west as illustrated in Figure G-8. The transmission lines would drop down to
the ground level and run along the west wall and feed directly into the EL. Note that eight TLs
would have a switch to redirect the power from the EL counter-ECCD entries to the eight UL coECCD entries of the upper steering mirrors of UL12 and UL13. The TL routing avoids supports tothe building structures to prevent additional distortions of the waveguide and the corresponding
mode conversion, as shown in Figure G-9.

Figure G-8: Space reservations for the EC upgrade with 2 possible options based on
PCR-496: 1) An extension to the RF Building (B15), as represented in the blue shaded area; or 2) In the
Assembly Hall with an area equivalent to the RF Building, but placed sufficiently far from the tokamak to
avoid any influence of stray magnetic field on the gyrotrons.

Figure G-9: (Left) Space reservations for the EC upgrade located in the back of the Assembly Hall with therequired space for 12 HVPS zones (2 levels), 21 RF sources (1 level) and associated TL. (Right) The routingof the TL would follow the west side of the building and be connected directly to the 24 entries of the EL.

G.4.1.2 Schedule of the 20 MW upgrade of the ECRH system
There are multiple procurement options as well as introduction of new technologies that could
simplify the schedule. However, the approach taken herein assumes the following:The design of the upgrade is made well in advance and is available prior to the decision date
(T0);

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The site infrastructure is installed starting at T0 and requires 18 months. This includes
modifications to buildings, cooling water (CCWS), Pulsed Power and Steady StateElectrical Networks (PPEN and SSEN), etc.;

The procurement and manufacturing of the first sets of HVPS and gyrotrons occur during
the 18 months of site preparation;

A single supplier provides the 12 HVPS sets and their delivery proceeds at a rate of
3 sets/year and an installation of two months. Thus, the full 12 HVPS are installed in
4.25 years;

Two gyrotron suppliers are used with a combined manufacturing rate of four tubes/year and6-month installation rate. Overlapping the installation and commissioning activities would
limit the 24 gyrotron installation to 6 years. This time period could be reduced pending theexperience gained from the baseline set of the first 24 gyrotrons;A single supplier is used for the 24 TL requiring 3 years for manufacturing and 1 year for
installation;Commissioning of the EC system up to the EL and UL entries requires 6 months.

The above is outlined in Figure G-10. The total time from T0 to be ready for commissioning withthe plasma is T0 + 6 years. Note that this assumes no interruption of the schedule due to ITER
operation, all available resources, two shift operation and pre-existing designs of the infrastructureand EC system upgrade layout.

Figure G-10: General schedule of the manufacturing and installation of the ECRH system upgradeof 20 MW (injected).

G.4.2 Upgrade of the ICRF system
The hardware for the ICH&CD system (PBS 51) is spread across building 11, 13 and 15 at the
ITER site, as shown in Figure G-11. In building 11, the ICRF system is installed in port cell 13 and15 at Level 1.
Overall, the ICRF system, including ancillaries, is installed in four buildings:


The RF Building hosts the HVPS and the RF sources (A);The Assembly Hall hosts the main transmission lines (B&D) and the matching system (C);369

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The Tokamak Building hosts some transmission lines (E), port cell prematching (F) andantenna (G);The Hot Cell Facility hosts the PPTF equipment.

Figure G-11: Layout of ICRF baseline system at the end of construction.

The baseline system has been designed to allow a future upgrade:

The antenna design is based on electric fields allowing operation up to 45 kV maximum
peak RF voltage;

TL and MS components are designed and tested for power levels corresponding to theupgrade power level.

As a consequence, the implementation of an upgrade of the ICRF system to 40 MW will not requirechanging or modifying any equipment in the Tokamak Building.
G.4.2.1 Description of the 20 MW upgrade for the ICRF system and its implementation
The upgrade of the ICRF system aims to increase the power coupled to the plasma from thebaseline level of 20 MW to 40 MW. All other parameters are unchanged: frequency range, spectra,
pulse duration, etc.
The upgrade will involve the following sub-systems:

HVPS: need to procure and install power supplies to feed 8 additional RF sources; nodevelopment needed; each unit identical to the ones being developed and tested for thebaseline. Switchgear and cables will be needed for connection to PPEN;

RF sources: need to procure and install 8 RF sources; these would be identical to the units
developed during construction. A prototype is being tested at present in IPR with successful
results;



TL: some components are needed to connect to main transmission lines; these components
will be standard components as used in the baseline configuration;370

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RF dummy loads and coaxial switches: a switching network on RF sources to be installed toallow testing and maintenance of sources; components are identical to the baseline
components installed in RF Building;Control system (data acquisition, safety, interlocks): procurement, software engineering,
installation and integration to interface the new equipment with the rest of the system;

Cooling system: The baseline RF power plant uses close to 300 kg/s cooling water from
CCWS2A; the upgrade will need a similar quantity; a provision for connection is available
in the Assembly Hall that can be used either for an ICRF upgrade or LHCD upgrade;

PPEN: the extension of the ICRF power plant will require the PPEN to supply an additional
50 MW; The connections are already planned in the PPEN network, but additional breakers
and distribution will be required;

SSEN: additional distribution required that could be taken from the existing infrastructureused for machine assembly in B13;

Building: the foreseen location of the upgrade is in the Assembly Hall, B13. Different
options are shown in PCR-496 [ITER_D_ETLLQC, 2013] and in Figure G-12. The
installation of the HVPS and RF sources will require erecting a three level metallic platform
in B13, similar to that described for the ECRH upgrade.

Figure G-12: Proposed location for the ICRF upgrade in North part of the Assembly Hall.

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G.4.2.2 Schedule for the 20 MW upgrade of the ICRF system
The schedule driver for the ICRF upgrade is the procurement of the RF sources estimated at 4
years. In parallel, the HVPS, TL and auxiliary equipment can be procured in coherence for a phased
installation. Final integration will require 1 year, after delivery of the last RF source.
The total duration must include contract preparation, hence the total time required for the upgrade isin the range of 6 years. Installation and integration takes place during the last three years (platform
erection, staged source & HVPS deliveries, etc.).
G.4.3 Neutral Beam Heating Upgrade
The Neutral Beam (NBI) system is spread across buildings 11, 34, and 37, area 30 and 31 at the
ITER site. In building 11, the Neutral Beam system is installed in the NB Cell (at L1, L2) and HV
Deck room (at L3) while the equipment in building 34, 37 and area 30 consists of NB high voltage
power supplies and the SF6 plant is installed in area 31.
The Neutral Beam installation is planned in two stages. The two HNBs and DNB are assembled in
Assembly Phases II and III, while a proposal for the installation of the captive components of the
3rd HNB is also under consideration and will be presented to the Change Control Board (CCB)
under a specific Project Change Request (PCR). Assembly Phases II and III are separated by the
HV tests, which are planned in parallel to the PFPO-1 campaign. The HV tests are an integrated testof the 1 MV HNB and 100 kV DNB power supplies through the transmission line to a spark gap
installed on the injector in the NB cell. The installation in the NB cell and the HV deck room at L3
in building 11 will be carried out in two shifts, while for building 34, 37, area 30 and 31 a single
shift is utilized.
G.4.3.1 Description of the NBI Upgrade – HNB3
The 3rd HNB injector, HNB3, is one of the proposed upgrades for the heating systems at ITER. The
Project Requirements [ITER_D_27ZRW8, 2014] prescribe 4 heating and current drive upgrade
scenarios to ensure the necessary flexibility for advanced operation regimes in ITER. The Project
Requirements also state that the ITER design shall not preclude the possibility of accommodating
the combination of heating and current drive upgrades, as specified in the Table 4-9 of the Project
Requirements (PR822-R). Based on these requirements the design of the HNB3 has been
considered in the design of the civil structures of the tokamak complex and the interfacing systems
needed to allow for this upgrade, when a decision is made on its need.
Unlike the upgrade options for the ICRF and ECRH systems, the HNB3 has strong implications andinterfaces with the Tokamak Building (B11). Upgrading the ECRH or ICRF systems would require
either an upgrade of the gyrotrons or RF sources, for ECRH or ICRF respectively, and theassociated power supplies, which are all located outside of B11. This greatly simplifies the upgradeof these systems and also allows for more flexibility as to when a decision can be made for theirimplementation.
The HNB3, due to its extensive interfacing and impact in the Tokamak Building, has a time
constraint, implying that the decision to install this equipment must be taken within a certain time
window. The NBI components are large and heavy components, (an injector is ~1200 t) and many
components can only be installed with hands-on access. This restricts the possibility of installationof HNB3 to the pre-nuclear phase. The baseline scenario is that the HNB3, if decided upon, would
be installed in Assembly Phase IV (pre-nuclear assembly phase). This installation is time
consuming and challenging; many plant systems will need to be dismantled in order to372

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accommodate the installation of the HNB3 and then these plant systems will be reinstalled uponcompletion of the HNB3 installation. Another challenge for the HNB3 is that, as for the otherinjectors, it is installed in the NB Cell. Access to the NB cell is limited and the baseline scenario toinstall the HNB3 requires dismantling of the West Gallery wall. This means that the NB Cell is
opened to the environment following operation with Be components in the tokamak, increasing therisk of Be contamination.
As the Staged Approach was developed, the understanding of the restrictions based on Be
contamination evolved further, resulting in the proposal to install the large components of the
HNB3 injector, which can only be installed by opening up the NB Cell to the environment prior tooperation with Be components in the plasma. This implies that the injector vessels (beam source
vessel and beam line vessel) and the passive magnetic shielding around it for HNB3 would beinstalled through the ‘north wall opening’ which is a temporary opening used to install the large
components of the HNB1&2 prior to PFPO-1. Once these components are installed, the north wall
can be closed resulting in confinement of the NB cell from the outside environment. Installing the
HNB3 components at the same time ensures that this confinement is maintained for the later stagesof the project.
If the large components of the HNB3 were not to be installed at this time, installation through thegallery is still possible; however, discussion with the nuclear regulator would be necessary to ensure
that, when considering the restrictions imposed to avoid Be contamination, this possibility would
still be realistic. This poses a significant risk to the potential upgrade of the HNB3.
The components of the HNB3 which should be anticipated for installation to ensure the future
upgrade of the NB system are detailed in the next section. When this proposed installation isformally accepted by the project, these components will become part of the construction Baseline.
G.4.3.2 Anticipated/captive components of HNB3
The large components of HNB3 are considered captive and it is proposed that they be installed in
Assembly Phase II. The components can be easily transported inside the NB cell through thetemporary opening in the north wall. The HNB3 components will be installed in parallel to DNB,
HNB1 & 2 components. The NB installation in building 11 will be carried out in two shifts.
The HNB3 captive components to be installed in Assembly Phase II are listed in Table G-2 .
Table G-2 – Captive/ anticipated components of the HNB3
Sr. No.

Description

Location

1

HNB3 Passive Magnetic Shield (PMS) base plates

2

HNB3 bottom Active Compensation and Correction NB cell
Coils (ACCCs) rails and retraction system

3

HNB3 PMS (except Top PMS#1 to 4b and High NB cell
Voltage Bushing (HVB) PMS)4

HNB3 Beam Line Vessel (BLV)

NB cell5

HNB3 Beam Source Vessel (BSV)

NB cell373

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G.4.3.3 Schedule for the NBI Upgrade – HNB3
All of the remaining HNB3 components (all except those considered captive and detailed in Table
G-2 above) that are required for the installation of a functional 3rd injector are considered as an
optional upgrade and a staged approach for installation is outlined in the schedule below.
As the HNBs at ITER will not operate before PFPO-2, a late decision based on their performance in
the ITER plasmas will exclude the possibility to have this upgrade in time for the pre-nuclear phaseinstallation, unless nuclear operations are delayed until this upgrade is implemented. However, theperformance of the beams will be well understood from the NBTF experiments, at latest by 2026,
and therefore a decision point could be considered at that time. Assuming that a decision is made in
2026, the estimated time to complete the procurement of the HNB3 can be seen in the schedule in
Figure G-13: .

Figure G-13: Schedule of activities for the HNB3 upgrade based on a 2026 decision point.

The latest possible time to make a decision on the HNB3 upgrade, to be ready to install a full HNB3
during the pre-nuclear shutdown, would be during the PFPO-1 plasma operation phase. By this time
both the performance of the beams at the NBTF and the performance of the other baseline heating
systems will be known, giving insight to determine the most effective upgrade path for the ITER
heating systems to ensure a successful implementation of the research program. Should such a
decision be made as late as the end of the PFPO-1 phase, it is still possible to install the HNB3
injectors for the DT phase. The corresponding schedule is shown in Figure G-14.

Figure G-14: Schedule of activities for the HNB3 upgrade based on an ‘end of PFPO-1’ decision point.

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G.4.4 The ITER Lower Hybrid System (LHCD)
The Lower Hybrid Heating and Current Drive (LHH&CD) system is not included in the ITER
construction Baseline. It is, however, considered as an H&CD upgrade scenario for the operational
phase. The LHH&CD system is designed to provide off-axis non-inductive current in any plasma
scenario and discharge phase. The system is therefore included in the Project Requirements as apossible heating and current drive system upgrade. The LHH&CD system is designed to deliver aminimum total Radio Frequency (RF) power of 20 MW. In some upgrade scenarios, the required
power level can reach 40 MW. The frequency foreseen is between 3.7 GHz and 5 GHz, with 5 GHz
agreed as the reference design frequency. The 3.7 GHz system could be a fall-back solution.
The 20 MW ITER LHH&CD system is composed of the following elements:

Wave Launcher: consists of four modules of radiating elements, arranged in two toroidal
and two poloidal rows, located in the ITER vacuum vessel via a port plug; these modules are
based on the passive-active multi-junction (PAM) concept;

Transmission Lines: transmit the RF power from the source to the launcher and transmission
components following paths of unequal electrical length;

The Power Sources together with their associated High Voltage Power Supply (HVPS),
auxiliaries and services: the operation of the whole system is monitored and co-ordinated by
an I&C system under the supervision of the ITER central control system (CODAC).

G.4.4.1 Status of the Lower Hybrid System design development
As this system is not considered part of the construction baseline, the main work has been to ensurethat the ITER facility can accommodate such an upgrade, should it be decided to design, construct
and install it at a later stage. In preparation for its possible integration, interfaces have been defined,
as far as possible, and requirements and interfaces have been documented:


System Requirement Definition [ITER_D_28B384, 2009];

Models physical implementation: space was reserved in the Assembly Hall for thegenerators and HVPS (Figure G-15); space allocations were reserved in the tokamak south
wall for penetrations; embedded plates are defined in the gallery and equatorial port cell 11;

Interface definition: the main interface sheets (with Tokamak Building, Assembly Hall,
CCWS, SSEN, etc.) have been developed and approved.

Figure G-15: Space allocation reserved in the Assembly Hall for the LHCD generators, transmitters andpower supplies.

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In parallel, the design of the system and the associated physics scenarios were reviewed from theinitial proposal within a task implemented by EFDA and involving all EU fusion institutions
interested in the LHCD system for ITER (WP09-HCD-03-01). The final report of this EFDA task
can be found in reference [ITER_D_SN28YH, 2016].
Despite large efforts to develop a klystron at 5 GHz (collaboration KSTAR-TOSHIBA) the tube
does not comply with all ITER requirements, in particular, the requirement on the Voltage Standing
Wave Ratio (VSWR). Indeed, due to the difficult coupling conditions for H-mode plasmas, thereflected power to the klystron may be high. For that purpose, the tube needs to deliver at least
500 kW at a (high) VSWR of 1.4. The main technical difficulty is linked to the requirement tooperate with very high reflected power while ensuring good efficiency [Ho 2011; Park 2013]. The
current performance is 500 kW, 600 s at VSWR < 1.2.
The remainder of the components of the LHCD system were also tested, but did not fulfil the full
specifications [Hillairet 2012; Hillairet 2015].
A backup solution at 3.7 GHz could potentially be considered:

Klystron TH2013C: CW, 620 kW at VSWR = 1.4, or 750 kW at VSWR = 1; demonstrated
successful operation at Tore Supra [Beunas, 2009];



PAM launcher: already tested and currently used on Tore Supra [Ekedahl, 2010; Dumont,2014];

Transmission Line: CW components currently used on Tore Supra [Goniche, 2014].

In addition, the conceptual launcher design would have to be adapted to the port plug environment,
which has changed substantially over the past years, with the implementation of PCR-439[ITER_D_ETJB4Q, 2014], and the changes that have been made to the bio-shield etc.
G.4.4.2 Lower Hybrid System port allocation and implementation issues
In the Project Requirements [ITER_D_27ZRW8, 2014], a number of ports were allocated to thepossible upgrade of the LHCD system, namely equatorial port 11, allocated to the launcher for the
upgrade scenario #1, ports 11 or 13 for upgrade scenario #2, and ports 13 (or 11) and 14 for upgrade
scenario #4. The various upgrade scenarios are described in Table 4-9 of the Project Requirements.
Note that, in the initial configuration, port 11 is a Diagnostics port, port 13 is an ICRF antenna port
and port 14 is an ECRH launcher port.
A number of studies were carried out, from the point of view of integration, and the preferred port
allocation was determined to be EP11. At present, the models have implemented in the tokamak
construction design for this port only [ITER_D_L9ABVC, 2014]. EP11 was considered the best
port due to its proximity to the Assembly Hall, where the transmitters are located, reducing the TL
length between the launcher and the transmitters. It should be noted that EP13 and EP15 would also
comply with this requirement. However, as the ICRF antennas are currently planned to be installedin EP13 and EP15, the magnetic connection between these systems could create perturbations on
the wave propagation from both systems. A conceptual study was also carried out on the possibility
to use EP8. This option was ruled out due integration and space reservation issues. Following thesestudies it was concluded that EP11 was the only viable option for the LHCD system in an upgrade
scenario.
In the initial configuration defined in the Project Requirements, EP11 is allocated as a Diagnostic
port and it is noted in the PR that: “Note 3: Diagnostics in port EP11 shall be removed for upgrade
scenario 1. In case diagnostics necessary for the machine protection and the basic plasma376

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operation are located in that port, they will be relocated in another port – or their functions will be
fulfilled by another diagnostic system. [PR1666-C]”
In PCR-738 [ITER_D_TL5BR3, 2016] it was finally decided to allocate the EP11 as a Diagnostic
port for the diagnostic systems shown in Table G-3.
Table G-3 – Diagnostics located in EP11 as per PCR-738
Diagnostic System

Functional from

55.QB: Eq Port #11 Systems

First Plasma

55.E3: VUV Survey

First Plasma

55.ED: X-Ray Crystal Spec Survey

First Plasma

55.B8: Activation System

PFPO-1

55.G4: Residual Gas Analyzers

PFPO-1

55.E2: H-Alpha (+ visible
spectroscopy)
55.EG: Divertor
VUV Spectroscopy

PFPO-1

55.E8: Neutral Particle Analyzer

PFPO-2PFPO-1

55.F2: Reflectometer (Main Plasma,
PFPO-2LFS)
The fully populated EP11 Port Plug and partly populated Interspace Support Structure (ISS) andPort Cell Support Structure (PCSS) will be installed at First Plasma. The ISS and PCSS contain
only components related to functional First Plasma diagnostics. The ISS and PCSS will be removed
after the First Plasma campaign and re-assembled to incorporate the components which were notpresent at First Plasma. The Port Plug may also be removed at this stage, if needed, to give
improved access to install in-vessel components. The complete and fully populated Port Plug, ISS
and PCSS will thus be installed during Assembly Phase II, ready for PFPO-1 operation. Unlessthere is a major failure, it is not expected that the EP11 Port Plug, ISS or PCSS should be removed
until after DT operations have started.
The decision to allocate this port to Diagnostics plays an important role in the realistic feasibility of
implementation of the LHCD upgrade. Other factors have to be also considered for such upgrade in
the context of the overall time schedule and a short summary of them is given below.
The projected cost of the LHCD system was developed under the EFDA task [ITER_D_SN28YH,
2016]. In this study the estimated cost for the full system is between €82 million and €87 million.
During this task is was also estimated that ~ 9 years, including the conceptual and detailed designphases, R&D, procurement, tests and installation, would be required to implement this upgrade.
This timeline was developed prior to understanding the complexity of design and integration into
the ITER environment for such a system and could therefore be considered optimistic. The upgrade
to include the LHCD system has a time constraint: if the LHCD system is not installed in the prenuclear assembly phase then it cannot be installed later. Even though the system can be designed tobe maintained by remote handling, installation would require human access in the Tokamak
Building. This implies that a decision on the need to implement such a system in ITER would haveto be taken by 2021 at the latest. In addition, the allocation of the EP11 to diagnostics systems in thebaseline approach means that no suitable port can be made available in this timescale.
The need for such an early decision on the LHCD system, prior to ITER operation, combined withthe unavailability of a viable port, raises significant issues concerning the possibility of maintaining
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the LHCD option as a viable upgrade path for ITER. To this end a review and assessment of the
H&CD needs for the development of the ITER long-pulse and steady-state scenarios has been
conducted in parallel with the development of the updated ITER Research Plan. This is described inAppendix E and concludes that upgrades of systems already included in the ITER Baseline (inparticular, NBI and ECRH) could provide similar functionalities to those expected from the LHCD
system for the steady-state scenario. Indeed, the most challenging feature of the Q = 5 steady-state
scenario is the achievement of very high confinement regimes in H-mode at high-q95 with H98 ≥ 1.6.
This is an essential requirement, which is independent of the H&CD upgrade chosen (i.e.
with/without LHCD) to achieve this scenario.

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Appendix H: Diagnostic Staging
H.1 Introduction
The diagnostic set on ITER is designed according to the requirements in the SRD-55[ITER_D_28B39L, 2017] and complemented by the applicable Project Requirements (PR)[ITER_D_27ZRW8, 2014].
To assess what parameters are measured and which system is measuring them, one can find thespecific parameter and the measurement range along with the diagnostic system that performs thatmeasurement in SRD-55 [ITER_D_28B39L, 2017]. The following is an example of the format forthe measurement of the core density profile:--- [55s1294]
Table MP054 – Measurement Parameter 054. Core ne shall be delivered with the following
specifications [55s1295]

The following diagnostic system(s) shall be provided to contribute to meeting this requirement: 55.C1 Thomson Scattering (Core) - Primary 55.F9 Reflectometer (Main Plasma, HFS) - Primary 55.C5 Toroidal Interferometer/Polarimeter - Supplementary 55.C6 Poloidal Polarimeter - Supplementary 55.F2 Reflectometer (Main Plasma, LFS) - Back-up
Where the word Primary means that the diagnostic is well suited to the measurement. Backup andSupplementary are more limited in what they provide for this measurement. Details of these can befound in the individual Diagnostic Design specific information. In general, all measurements haveat least one Primary system to provide them; often, they are dependent on one or more
Supplementary contributions for performance and/or a Backup contribution for reliability oravailability reasons. All the measurements planned and the corresponding systems can be traced
through the SRD-55 [ITER_D_28B39L, 2017] above.
The recent Staged Approach, with four operational phases (including First Plasma), has defined in
the project when each system can be available. These will be detailed further below where eachphase will be discussed in turn identifying the status of the different systems. For First Plasma,
there are some cases where temporary systems are installed to cover the fact that the more advanced
system will arrive in phase II. This will be explained below.

H.2 First Plasma Diagnostic Capabilities
For First Plasma, the set of systems that will be available is shown in Table H-1 . It can be seen thatthese cover mainly the basic systems that are needed for establishing the main parameters of theFirst Plasma. In this phase the emphasis on measurement capability is for magnetics, plasma
breakdown, investment protection (hard X-rays) and density for a plasma with a current of up to
1 MA (presently considered for the Engineering Operations phase after First Plasma).

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Table H-1 – First Plasma and Engineering Operations diagnostic systems
PBS

System Name

Comment

55.A0

Magnetics System Electronics & Software

55.A1
55.A3/A4/A9
55.A5/A6
55.A7/AD/AE/AI
55.AA/AB/AC
55.AF/AG/AH
55.E2
55.E3

Continuous External Rogowski
Outer Vessel Coils
Steady State Sensors
Flux Loops
Inner Vessel Coils
Diamagnetic Sensors
H-alpha/Vis in EPP12
Vacuum Ultra-Violet Survey
Visible Spectroscopy Reference System Temporary version, as(partial)
port plug not installed
X-Ray Crystal Spectrometer
Hard X-ray Monitor
Only for FP
Vis/IR Equatorial in EP16 (temporary)
1 out of 3 views fully
Vis/IR Equatorial in EPP12 (partial)
working
Temporary for FP
Pressure Gauges (temporary)
Stray ECRH detector
Temporary for
TF Mapping (one-off activity)alignment of theblankets before FP.
In-Vessel Lighting
Tokamak Structural Monitoring system
Single channel
Density Interferometer (partial)
installed without port
plug

55.E6
55.ED
55.EE
55.G1.E0
55.G1
55.G3.50
55.GB
55.GF
55.GL
55.GT
55.FA

A sufficient set of the magnetic sensors will be functionally available; most others will be there
either trapped or to benefit from joint commissioning with the PF coils and TF Coils. This phasewill include an extensive commissioning phase, in advance, to ensure all the systems are ready.
Where not stated otherwise, these systems will be fully installed with the exception of what is called
the temporary sub-systems or partially installed systems. These are:55.E6 Visible Spectroscopy, which will have a single radial channel for visible light observationthat will be reconfigured after Phase I.
55.G1.E0 Infrared system, which is a special Infrared system to observe the special ECRH mirror
needed for first plasma. This will be removed after FP.
55.G1 Visible/IR system is the conventional system for Vis/IR measurements but only one channel
will be functional.

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55.GL In-vessel Lighting system is (part of) the in-vessel lighting system to allow visual inspectionof the in-vessel surfaces with the cameras of 55.G1
55.G3.50 Pressure gauges will include fast pressure gauges set up close to the back of an empty port
plug. These will have a fast time response (better than 100 ms) sufficient for monitoring the
breakdown of the plasma.
55.GF Is a temporary system to map the toroidal field to confirm/improve FW alignment.
55.FA Interferometer (DIP), which will have a radial view of a retroreflector on the inner wall and,hence, provide a line averaged density measurement. It will be reconfigured after First Plasma andEngineering Operations.

H.3 Diagnostic Capabilities for PFPO-1 and planned upgrades
As PFPO-1 is a much more demanding phase, the number of systems installed is increased
significantly compared to First Plasma. This can be seen in Table H-2. In this table the systems inblack are expected to be fully operational after a suitable commissioning phase. In this phase ofoperations the emphasis on measurement capability is for plasma control, investment protection andbasic plasma parameters. Note that, as for the First Plasma systems, the measurement capabilities
for each plasma parameter can be found in SRD-55 [ITER_D_28B39L, 2017]. A discussion on theimplication of 1.8 T operation for the diagnostic systems available in PFPO-1 can be found insection H.6.
To explain the state of those systems that are not complete for PFPO-1 operation (blue for upgrades
and red for enabled systems in Table H-3 ), a short explanation is provided below:Fibre Optic Current Sensor (55.A8 FOCS)

The Fibre Optic Current Sensor (FOCS) system will use optical fibres placed around the Vacuum
Vessel (VV) to measure the plasma current by means of the Faraday Effect. The optical fibres will
be placed inside stainless steel tubes which protect the fibres from mechanical damage and allow
fibre replacement. These tubes will be fixed with clamps on the outer VV surface. The tubes
themselves will provide the boundary between vacuum and non-vacuum.
The FOCS presents an attractive alternative to other steady-state plasma current diagnostics. Its
operation is based on the detection of the Faraday rotation experienced by a polarized light in an
optical fibre because of the presence of magnetic field aligned with the fibre axis. The complete
sensor requires electronics for the signal processing, but only the sensing fibre is placed in thevicinity of the burning plasma (around the ITER VV) and no front-end electronics, which could be
subject to the ITER harsh environment, are required. It was also demonstrated that the radiationinduced absorption levels in radiation hard optical fibres are compatible with the requirements for
signal transmission for the full ITER lifetime. The current FOCS design with the optical fibres
inserted into stainless steel tubes using a blow-in technique means the sensing head can be easily
replaced in case of degradation due to service conditions or development of an improved sensing
technology. Furthermore, the optical fibre which is used for the FOCS allows, in principle,
distributed high-resolution measurements of other parameters, e.g. temperature, strain, or radiation
dose. A distributed magnetic field measurement using the FOCS structure also seems possible. The
FOCS installation can potentially host other systems, for example an activation wire for neutron
dosimetry.
The system is enabled only: Components up to the cryostat (Figure H-1) will be installed but not
functionally tested. Fibre bundles through the building penetration will also be installed. Fibre joints
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within port cells and diagnostic hall may or may not be installed depending on cabling schedule andfunds. The remainder of the system (bundles in the port cell, polarimetric system in the diagnostichall) will be installed if funded.
In Vessel calibration (55.BV)-(PFPO1-PFPO2-FPO)

The total neutron yield (Yn) measurement is the direct measurement of fusion power and it has to beprovided by neutron diagnostics with an accuracy of 10% and 1 ms time resolution. This is a strict
requirement as it relates to knowing the fusion power generated in ITER. To ensure theserequirements the neutron diagnostics have to be calibrated and a strategy has been developed. It
consists of three different stages – (a) first is the partial in-vessel neutron calibration campaign withlimited toroidal/poloidal scan performed before PFPO-1, (b) second is the full toroidal and poloidal
scan at 2.5 MeV neutron energy performed before PFPO-2 and (c) the third is again full scan but at
14 MeV neutron energy performed before FPO. The objective of these campaigns is the comparisonof the experimental data/sensitivities with the neutron transport modelling, determining thecalibration factors and verifying of the ITER neutronics model.
In the In-vessel neutron calibration campaigns, a neutron source (Californium-252 or DD/DT sealed
neutron generators) will be moved inside the vacuum vessel in different toroidal and poloidal
positions irradiating the most sensitive detectors (55.B4 NFM in EQ#1, 55.B8 NAS- Neutron
Activation System in Upp 18 and 55.BC DNFM- Divertor Neutron Flux Monitor in Lower Port 2,
Divertor cassette 3). Investigation is going on the possible use of RH transportation systems or by amanually handled system (pneumatic piping, train transporter or staging) to perform these
movements.
In the near future, the technical specification will be defined for neutron sources (Californium-252)
and, in particular, for sealed neutron DD/DT generators.

Figure H-1: Overview of the FOCS integration in ITER

Neutron Storage Area (55.BT)

The radiation sources for neutron detector testing and In-Vessel neutron calibrations as well thefission chambers (FC) of the 55.BX Neutron diagnostics need a storage area on the ITER site. The
neutron diagnostics (55.BC DNFM, 55.B4 NFM, 55.B1 RNC, 55.B2 VNC, 55.B3 MFC) contain382

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fission chambers (with different amounts of Uranium-235 and Uranium-238) and will be delivered
to ITER starting in the 2020s from Japan Domestic Agency (JA-DA), Russia Domestic Agency(RF-DA), China Domestic Agency (CN-DA) and EU Domestic Agency (EU-DA). The fission
chambers have to be stored at ITER site before they are assembled/ installed in the machine. The
diagnostics will also use radioactive sources such as ,  and neutron sources for calibration, toperform site acceptance test and functional tests.
Meetings have been held with Facilities and Logistics Management division and safety and security
officers within the IO. The storage location has to ensure safety and security of the fissile andradioactive material under normal and off-normal conditions, and has to be inside the ITER INB
zone. The IO has to make a declaration to the ASN in order to have Uranium-235 on site andrequest authorization to ASN for radiation sources on site. The Domestic Agencies will also have toapply for an Export Licence to their respective governments and the whole life cycle of the fission
chambers has to be known including the activities in the storage area.
Currently, a storage area has been enabled in the baseline. A layout is being designed representing
the necessary required storage for fission chambers and radiation sources. In the storage facility
there will be a dedicated area for performing the functional tests of the detectors by using very low
yield radiation sources.


H-alpha and Visible Spectroscopy (55.E2)

The H-alpha and Visible Spectroscopy diagnostic on ITER shall consist of 4 optical channels
located in several port plugs: two channels in EPP#11, one channel EPP#12 and one channel in
UPP#2. The wide field of view (FOV) channels located in EPPs (two channels in EPP#11 and one
in EPP#12) shall be aimed to observe the whole poloidal section of ITER FW (at least half aBlanket Module wide in toroidal direction). The EPP#11 channels shall cover the inner wall and top
part of the chamber, whereas EPP#12 channel shall cover most of the ITER FW outer part (BM#10BM#18). The approximate FW areas to be observed by EPP channels are shown in Figure H-2.

Figure H-2: Fields of views of H-alpha and Visible Spectroscopy.

Strong impurity emission from wall materials (e.g. Be) can indicate excessive levels of interaction
between the plasma and the wall and indicate regions of potential erosion and redeposition ofmaterial leading to consequent tritium retention and dust production. Therefore the need of Be
imaging was identified instead of the currently specified integral measurements. For the same
reason imaging of the D and T influx was requested.
The design of the diagnostic allows for imaging, however this option is not in the baseline and thus
needs to be an upgrade. Extra cameras are needed to realize this; the optical resolution of the system
is about 15 mm on the inner wall and 45mm on the outer wall.
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Divertor Impurity Monitor (55.E4)

The baseline for the Divertor Impurity Monitor (DIM) system has 6 views x 70 sightlines per view.
This system allows the diagnostic to perform the Machine Protection role of measuring the impurity
emission. In this configuration, roughly half of the sightlines are populated with detectors andspectrometers. To improve the spatial resolution for the ion temperature measurement (fromapproximately 40 mm to approximately 20 mm), more spectrometers are required. Note that theformal requirement is actually 50 mm (along leg) x 3 mm (across leg). Achieving 3 mm is very
challenging even after data unfolding, but populating all the sightlines would be significant step
closer to the objective and this is included as an upgrade.
Radial X-ray System (55.E7)

The radial X-ray system comprises of in-port and ex-port detectors. In the original plan, it was
assumed that the initial non-nuclear detectors would be replaced by nuclear detectors before thestart of the DT phase. However, the current plan is that the port plugs will only be removed fromthe machine and opened at the end of the First Plasma Phase (just before PFPO-1). This means thatthe detectors need to be developed more quickly than previously planned. As a result of theconfiguration workshop, funds have been put aside for the development of these detectors and wehave worked with the DA involved to reorganize the planning. This work will need to be completed
by the time the port plug (EQ12) comes out for complete assembly after the first plasma campaign.
Density Interferometer Polarimeter (55.FA)

The Density Interferometer Polarimeter (DIP in Eq port 8) is of the dispersion interferometer type
and will comprise of two channels. One which goes radially inwards and one which goestangentially from Equatorial port 8 to Equatorial port 3. The tangential chord supports thepolarimeter. Following the basic single channel radial installation in the First Plasma phase, thesystem will be reconfigured and expanded to its planned capacity. In this case, all of thecomponents in port 8 will be installed and roughly half the back-end. In the current planning, thebudget for the completion of this system will be available in the operational phase and thus is an
upgrade. Note that the TIP (Toroidal Interferometer Polarimeter) in Equatorial 9 is also planned tocome on line at this stage. These systems will operate independently.


Flow Monitor (55.GE)

The ITER boundary coherence imaging (CI) system, or simply ‘flow monitor’, is an advanced
imaging system for detailed measurement of the impurity ion flows and temperatures in the ITER
plasma boundary and divertor. These measurements are essential for benchmarking models of edge
and divertor transport. Moreover, the layered co-deposition of beryllium and isotopes of hydrogenin the ITER divertor is a safety issue because of possible tritium retention and the production of
dust. A measurement of ion flows between the inner wall and the inner divertor baffle, combinedwith erosion/deposition measurements and Langmuir probe measurements at the inner baffle willallow validation of Be migration models. Validated models will provide a predictive capability fordivertor performance and control in ITER and future fusion devices. The schematic of the ITER
flow monitor is shown in Figure H-3. The diagnostic consists of an optical system with a front-end
mirror system located inside one of the ITER port plugs and extending through interspace to theport cell which contains the processing optics and CCD camera. The flow monitor is planned for
installation in equatorial port 8. A scoping study for the ITER CI flow monitor has been performed
and thus this system is currently enabled.

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Core Thomson Scattering (55.C1)

The system will bet set up initially with a single well established laser (1064 nm). This will
typically have a frequency of between 20 and 50 Hz. The second or further lasers to make up to
100 Hz will come in PFPO-2.
This system will guarantee electron temperature measurements to about 25 keV within
specification. For higher temperatures as in the requirements (up to 40 keV), several additional
measures are being considered. One is an additional longer wavelength (1320 nm) laser and theother is to use polarization scattering on the conventional laser. Both methods use the sametransmission chain.

Figure H-3: Outline of the Divertor Flow Monitor.



Residual Gas Analyzer (55.G4)

The Residual Gas Analyzer (RGA) (see Figure H-4) measures the neutral gas composition in adivertor port and an equatorial port during plasma operation. The 2 RGA subsystems are being
procured by the US-DA and are planned for this phase. Only vacuum flanges are delivered for theFirst Plasma to blank-off the RGA sampling points.
Each subsystem comprises a sampling pipe which feeds gas from the primary vacuum at the port toan analysis chamber located in the port cell. In the divertor port the RGA is only able to achieve therequired response time, 1 s, when there is a gas flow to the divertor cryopump; typically thesepumps are on for only 50% of the plasma discharge. The upgrade of the RGA beyond the scope of
the PA with the US-DA consists of a further 2 RGA systems at 2 other divertor port locations, sothat continuous monitoring of the divertor is possible with suitable phasing for cryopumps.

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Figure H-4: RGA integrated into the Equatorial 11 Port Cell Support Structure.

Table H-2 – Diagnostic systems available for PFPO-1 (in addition to those for First Plasma)
PBS
55.A8
55.AJ
55.AL/AO
55.AM
55.AN/AP
55.B3
55.B4
55.B8
55.BC

System Name
Fibre Optic Current Sensor
High Frequency Sensors
Divertor Coils
Divertor Shunts
Divertor & Blanket Rogowski Coils
Microfission Chambers
Neutron Flux Monitor Systems
Neutron Activation System
Divertor Neutron Flux Monitors

55.BV

Neutron Calibration (2.5 MeV) [Partial]

55.BT

Neutron Facility Area

55.C1

Core Plasma Thomson Scattering (partial)

55.D1

Toroidal Interferometer polarimeter (TIP and measuresline integrated Density in EQ9)
Bolometry System

55.E2

H-Alpha Visible

55.E4

Divertor Impurity Monitor

55.E5
55.E6

Core Imaging X-ray Spectrometer
Visible Spectroscopy Reference System

55.E7

Radial X-Ray Camera

55.EG
55.EH

VUV Divertor
Vacuum Ultra-Violet Edge

55.C5

Comment
Captive part installed

Simple calibration testplanned here
1 laser installed in thisphaseSome cameras missing
at this stage.
Some
spectrometers
missing at this stage.

Edge Detectors still
under development386

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55.EI
55.F1

X-Ray Crystal Spectroscopy Edge
Electron Cyclotron Emission

55.FA

DIP (density-EQ8)

55.G1
55.G2
55.G3
55.G4
55.G6
55.G7
55.GA
55.GD
55.GE
55.GG

Vis/IR Eq Ports (remainder of system)
Thermocouples (divertor)
Pressure Gauges
Residual Gas Analyzers
Divertor IR Thermography
Langmuir Probes
Vis/IR Upper Ports
First Wall Samples
Flow Monitor
Calorimetry (for testing with aux. heating inputs)

All front-ends will be
installed.

2 channels installed

Captive parts installed

H.4 Diagnostic Capabilities for PFPO-2 and planned upgrades
By PFPO-2, a strong diagnostic set will be established with many systems already in operation
since PFPO-1. In this phase, the emphasis on measurement capability is for plasma control,
investment protection and basic plasma parameters. The diagnostic set added or upgraded in thisphase is shown in Table H-3 . Note that, as for the First Plasma systems, the measurement
capabilities for each plasma parameter can be found in SRD-55 [ITER_D_28B39L, 2017]. A
discussion on the implication of 1.8 T operation for the diagnostic systems available in PFPO-2 canbe found in section H.6. From an operation perspective, this is a much more demanding phase, as itis the phase in which 15 MA/5.3 T plasmas will first be achieved (see section 2.5.5.11) and all
systems will be expected to be highly performing in preparation for the transition to D/DT
operation. This includes, in particular, the systems that are brought into routine operation in thisphase. Note that many of these PFPO-2 systems will already be installed in PFPO-1. For them basic
commissioning will be attempted, where possible, using PFPO-1 plasmas.
To explain the state of those systems that are not complete for PFPO-2 operation (blue for upgrades
and red for enabled systems in Table H-3 ), a short explanation is provided below:Edge Thomson Scattering (55.C2)

The edge Thomson scattering will have most of its components installed in PFPO-1. This will allow
it to come in to operation reliably in PFPO-2 with the possibility to commission it in the second halfof PFPO-1, if opportunities are available. If this is possible, then this diagnostic can play a role inmeasuring the edge profiles for the 1.8 T campaign; this is of great interest for the characterizationof the H-mode pedestal plasmas in 5 MA/1.8 T plasmas in PFPO-1 (see section 2.5.4.9). The
system was originally designed with a spatial range of measurement of 0.85 < r/a < 1.0. To allow an
overlap with the core Thomson Scattering, the viewing range is being extended to 0.8 < r/a < 1.03.
This will also allow having measurements for a wider range of plasma shapes/positions. Therefore,
the system is currently being designed with this capability, which is included as an upgrade. The
main items of the upgrade are the additional fibres and back-end electronics for the new channels.

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Poloidal Polarimeter (PoPola) (55.C6)

The PoPola system measures the plasma current profile by measuring the change in the polarization
properties of a fast infrared (FIR) laser beam traversing the plasma, due to the magnetic field.
Tomographic reconstruction on an array of laser chords spanning the plasma yields the 1-D
q-profile. As with all tomography techniques, increasing the number of chords increases accuracy
and reduces errors. The baseline PoPola system has 10 chords, and ‘enables’ an additional 3 chords
by providing front-end (in-vessel) components. The back-end (optics relay, lasers and detectors) for
the 3 additional chords is considered an upgrade. As shown in Figure H-5, the average error acrossthe profile is reduced with these additional 3 chords and, therefore, this upgrade is adopted in the
diagnostic plan.
High Temperature Thomson Scattering (55.C8)

Accessing the highest temperatures expected in ITER (up to 40 keV) put strong demands on theThomson Scattering measurement technique. This comes about because at temperatures above 25 to
30 keV, the spectrum is very much shifted towards the blue spectrum. At these wavelengths, there
are some issues related to background lines and general detectability of the signals. If the coresystem can solve these issues (see 55.C1 for details), then this system will not be needed. A
decision on this system will need to be taken in the next two years and work is ongoing to support
this decision. It should be noted that integration of this option is complex.
a)b)

Figure H-5: a) 10 PoPola chord layout and q reconstruction errors, b) 13 chord PoPola layout and q
reconstruction errors. Average error across the profile are reduced from ~15% to <10%.

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Laser Induced Fluorescence (55.EA)

The LIF (Laser Induced Fluorescence) diagnostic is intended to support the measurement of thehelium density in the divertor. This system is currently enabled and is sharing the front-end with theDivertor Thomson Scattering system. Interfaces are preserved for this system.


Motional Stark Effect (55.EB)

The MSE (Motional Stark Effect) diagnostic measures the local magnetic field strength anddirection based on the wavelength split and polarization of the Stark split line emission of neutralbeam atoms. From these measurements the plasma current profile (q-profile) is derived.
The baseline MSE system on ITER has one view on Heating Neutral Beam 1 (HNB1) fromequatorial port 1 (EPP1), measuring the plasma centre, and one view on HNB2 from EPP3,
measuring the plasma edge (see Figure H-6). Because the injection direction of the HNBs can bevaried between an on-axis (HNB passing through the plasma centre) and an off-axis position (HNB
passing below the plasma centre), no measurement of the q-profile can be provided in the plasma
centre when HNB1 is aimed off-axis. This is problematic because the main aim of off-axis HNB
operation is to create flat or reversed q-profiles, which are difficult to determine experimentally
without an on-axis q-profile measurement.
As a solution, the front-end of the MSE diagnostic has been modified to include a view on the
diagnostic neutral beam (DNB) from EPP3 (see Figure H-6). Although the DNB has a more limited
duty cycle and lower signal levels than the HNBs, it is always aimed on-axis. This leads tofollowing MSE capabilities for different modes of HNB operation:


HNB1
HNB2

on-axisoff-axis&

Full (core & edge) coverage by the MSE system.
Core is covered twice (HNB1 and DNB), allowing tocommission the DNB view which will have a lower
signal.



HNB1
HNB2

off-axis
off-axis&

Full (core & edge) coverage by the MSE system.
Core is covered by the DNB view.



HNB1
HNB2

on-axison-axis&Core MSE coverage, but no edge coverage.
Acceptable because the edge q-profile is well coveredby magnetics and polarimeter.



HNB1
HNB2

off-axis
on-axis&Core MSE coverage by DNB view, but no edge
coverage. This HNB configuration is, in principle,
identical to HNB1 on-axis and HNB2 off-axis, which
has the best MSE coverage.

The baseline MSE does include the front-end (in-vacuum) optics for the DNB view, but an upgrade
is required for the (in-air) back-end; this includes optical fibres (most critical, as they have to beinstalled no later than Assembly Phase II before PFPO-1), detector systems and instrumentation andcontrol (I&C).


BES fluctuations and Escaping Alpha FICX (55.EC)

Beam Emission Spectroscopy (BES) fluctuation measurements are used in many tokamaks to study
both MHD and turbulence, typically using a 2-D array coverage that allows identifying radial andpoloidal correlation lengths, EB velocities and shearing. Especially in the pedestal region, BES
fluctuations are of interest in the study of the L-H transition and ELM characteristics.
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Figure H-6: Top-view and cross section view of the MSE sightlines. The baseline MSE system covers the
on-axis HNB1 from EPP1 (core measurement) – green – and the off-axis HNB2 from EPP3 (edgemeasurement) - blue. It also includes the front-end optics (in-vacuum mirrors) for the DNB view – black –but not the optical fibres nor the back-end detection system and I&C.

Fast -particles escaping the plasma before thermalizing are a concern, because they create an extra
heat load on the walls and they reduce the self-heating of the plasma. Measuring/monitoring themis, however, quite challenging. Fast Ion Charge Exchange (FICX 7) is predicted to have a substantial
signal from the pedestal region upon interaction with HNB2 for escaping- fractions down to 10%.
7

FICX is in current tokamaks mainly used to study fast neutral beam particles and therefore named FIDA (Fast Ion DAlpha). In ITER, however, the main fast ions of interest would the -particles, rather than the D from the beams, hence
the use of the more general term Fast Ion Charge Exchange or FICX.

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This diagnostic would monitor the quantity of escaping -particles and also yield information about
the energy distribution of the escaping ’s.

Figure H-7: Sketch of the proposed upgrade that would allow both BES fluctuation and escaping- FICX
measurements. The front-end (in-vacuum) optics already form part of the CXRS Edge system. The upgrade
would involve the installation of dedicated optical fibres, back-end detection systems and I&C. The inset
gives an overview of the proposed usage of the CXRS Edge image plane, whereby a 15 by 10 array of BES
fibres (blue) would be interleaved between the on-axis (active) CXRS channels (dark green) and 3 FICX
channels (red) would be interleaved between the off-axis (passive) CXRS channels (light green).

Neither BES fluctuations nor FICX measurements currently form part of the ITER baseline.
However, the neutral beams (DNB and HNB2) are already covered in the pedestal area, withadequate spatial resolution, by the CXRS (Charge Exchange Recombination Spectroscopy) Edge
system. Therefore, an upgrade to enable BES fluctuation and escaping- FICX measurements
would only involve the (in-air) back-end: this includes optical fibres (most critical as they have to
be installed no later than Assembly Phase II before PFPO-1), detector systems and I&C. Figure H-7
shows a sketch of the proposed system, whereby a 15 (radial) by 10 (vertical) array of BES
channels is interleaved between the on-axis (active) CXRS channels and 3 FICX channels (thatrequire more fibres to collect more light than standard CXRS channels) are interleaved between the
4 off-axis (passive) CXRS channels. In the diagnostic area the BES fibres would be connected to a391

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fast, filtered detector array, whereas the FICX fibres would be connected to off-the-shelf highétendue spectrometers.
Table H-3 – Diagnostic systems available for PFPO-2 (in addition to those for First Plasma
and PFPO-1)
PBS
55.BV
55.C1

System Name
Neutron Calibration (2.5 MeV)
Core Plasma Thomson Scattering (full)

55.C2

Edge Thomson Scattering

55.C4
55.C6

Divertor Thomson Scattering
Poloidal Polarimeter

55.C8

High Temperature Thomson Scattering

55.E1

Charge Exchange RS Core

55.E8

Neutral Particle Analyzers

55.EA

Laser-Induced Fluorescence

55.EB

Motional Stark Effect

55.EC

Charge Exchange RS Edge

55.EF
55.F2
55.F3
55.F9
55.G8
55.G9
55.GG

Charge Exchange RS Pedestal
Reflectometry Low Field Side
Plasma Position Reflectometry
Reflectometry High Field Side
Erosion Monitor
Dust Monitor
Calorimetry

55.GC

Tritium Monitor

Comment

r/a from 0.85 to 1.0
available
10 channels available
Still to be decided if this isneeded as the Core system
may be able to handle thehigh electron temperature
Requires diagnostic neutralbeam to be operational
Enabled and used Divertor
Thomson Scattering frontend
Requires heating neutral
beams.
Requires extra components
and diagnostic neutralbeam to measure when
heating beam is off-axis.
Requires diagnostic
neutron beam.
Requires extra components
to add measurements offluctuations and escaping
’s.

Commission in PFPO-2
and obtain first results392

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H.5 Diagnostic Capabilities for FPO and planned upgrades
By FPO, many diagnostics will be established with many systems already in operation since First
Plasma, PFPO-1 and PFPO-2. In this phase the emphasis on measurement capability is for fusion
power production and burning plasma physics. The diagnostic set added or upgraded in this phase isshown in Table H-4 . It can be seen that the systems still to be implemented are those that are more
specifically related to DT measurements. All these systems are currently having their interfaces
defined and the basic system is designed. Most of these systems will undergo a Conceptual Design
Review) CDR in the near future year with the exception of the CTS which has already been
completed.


Radial Gamma Ray Spectrometer (55.B7)

The Radial Gamma-Ray Spectrometer (RGRS) is a multiple detector, multiple lines-of-sight
spectrometer optimized for photons in the energy range up to 17 MeV.
During fusion discharges in ITER, a series of energetic gamma-ray lines are expected, of which themost interesting ones are at 3.21 and 4.44 MeV (from α-Be reactions) and 17 MeV (from D-T
reactions). These gamma lines are expected to be used as a sensitive diagnostic for the confined
α-particle distribution inside the plasma volume and, secondarily, to evaluate the fusion reaction
rate. The confined -density profile may be inferred from the measured intensity of the 4.44 MeV
gamma ray emission from the 9Be(α,nγ)12C reaction. By measuring this emission along several linesof sight, the profile can be obtained through standard inversion algorithms. Moreover, RGRS is
expected to have a prominent role in the diagnosis of runaway electrons (REs), particularly while
they are in plasma (before hitting the wall); their maximum energy and other important parameters
can be derived from the gamma measurements.
In order to exploit the potential of such a diagnostic, it is necessary to measure: i) the intensity(count rate), ii) the spectrum and iii) the distribution of the emitted photons, and all with anadequate time resolution in order to follow the plasma evolution. To this aim, a multiple lines-ofsight spectrometer has been designed, which matches the requested goals of energy resolution,
efficiency and counting rate capabilities; the latter are needed to follow the time evolution of rapid
phenomena in plasma discharges. The multiple lines-of-sight approach allows the reconstruction of
the emission profile inside the plasma volume through a tomographic reconstruction approach.
The main components of a single RGRS detector are: i) the LaBr3 scintillator crystal, ii) thephotomultiplier tube (PMT), and iii) the magnetic shielding for the PMT. The detector assemblies
for RGRS are designed to give support for the detectors while also being a component of themagnetic shielding. These are made of two main parts: a detector container/magnetic shield and an
assembly support. The RGRS will perform gamma-ray measurements along a view of thecollimated channels of the ITER Radial Neutron Camera (RNC) by the design of a suitable set ofdetectors, collimators and attenuators. This system is currently enabled and a CDR is expected in
the near future.


Lost Alpha Monitor (55.B9)

In ITER, the fusion reactions and the use of neutral beam and ion cyclotron heating systems will
generate fast ions. Fast ions can be expelled from core region to the plasma edge bymagnetohydrodynamic (MHD) perturbations. Fast ion loss can adversely affect confinement andplasma heating and can lead to localized heat load on the plasma-facing components.

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The fast ion loss detector (FILD) (see Figure H-8) is one of the most widely used diagnostics formeasuring fast ions in the plasma edge. FILD provides the measurement of the fast ion energy andthe velocity pitch with time response up to the Alfvénic frequency, which allows extracting
important information about the underlying fast ion loss processes. FILDs are installed in all major
tokamaks and this diagnostic has been identified as a priority fast ion diagnostic for ITER by the
ITPA Energetic Particle Topical Group [ITER_D_UFAXKQ, 2017].
A scoping study for the ITER FILD, slated for installation in equatorial port 8, is currently being
performed. The conceptual design consists of a reciprocating system containing a probe headequipped with a scintillator (combined with an array of the Faraday cups) and an optical systemtransferring the signal from the scintillator to an array of photomultipliers and to a CCD camera.
In addition to the FILD, a lost- monitor system has been proposed for installation under thedivertor dome. This system will consist of an array of Faraday cups and will complement FILD
measurements by measuring marginally trapped fast ions in continuous data acquisition mode.

Figure H-8: FILD system outline as conceived for Equatorial 8.

High Resolution Neutron Spectrometer (55.BB)

The High Resolution Neutron Spectrometer (HRNS) is dedicated to measure time-resolved neutron
spectra for both DD and DT plasmas, providing mainly the determination of the fuel ion ratio in theplasma core for a wide range of ITER operational scenarios in fusion power (neutron yield)
including spectroscopy measurements in the initial deuterium phase. The supplementary functionsof HRNS are to provide information on the ion temperature, confined -particles and fast ions.
HRNS is a one-channel collimator system and is able to perform neutron measurements providing
line-integrated data, where the record signal is an integral over local plasma conditions along the
instrument’s line-of-sight (LOS). The HRNS will be installed is in the Equatorial Port Cell #1
behind the RGRS and the RNC.
The HRNS system provides a dedicated measurement of the 14 MeV neutron emission from ITER
DT plasmas and also of the 2.45 MeV emission in the initial phase of D operations. A suitable394

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2.45 MeV spectrometer is an important part of the HRNS system, which can provide important
information during scenario development and testing of the heating systems in D. In addition, due
to the long pulses and good confinement properties of ITER D H-mode plasmas, the tritium
inventory in D plasmas may be sufficiently high to provide a good opportunity for testing andtuning all of ITER’s 14 MeV neutron diagnostics by the burn-up of the T produced.
The neutron spectrometers should provide measurements of nT/nD with high time resolution(100 ms) and a measurement uncertainty of 20%. Fulfilling these requirements cannot be achieved
with a single neutron spectrometer. Most neutron spectrometers have a dynamic range of 3 to 7 due
to detection efficiency and/or count rate capability limits, set by random background and/or pile-up.
This implies that the HRNS system needs to be built up of a number of different spectrometer
techniques. For high performance plasmas generating a high neutron flux at the HRNS position,
spectrometers based on the thin-foil proton recoil technique is preferred. For the medium range ofneutron fluxes two spectrometer types can be used: back scattering time-of-flight and diamonddetectors. For low nT/nD levels and pure D plasmas the only option is a conventional ToF system.
The HRNS is an enabled system and is planned to undergo a CDR in the near future.
Vertical Gamma Ray Spectrometer (55.BD)

The Vertical Gamma Ray Spectrometer (VGRS) system is currently foreseen to fit in the upper port
in the place of the Vertical Neutron Camera (VNC). This system is enabled and the ability to install
it is maintained by defining its interfaces. Currently in ITER, it is not possible to have both the
VNC and the VGRS together due to space limitations, thus the priority has been given todeveloping the VNC in the first instance. Note also that background development is needed toprovide a reliable system for the vertical views.
Tangential Neutron Spectrometer (55.BE)

The Tangential Neutron Spectrometer (TNS) is a diamond-detector-based spectrometer with
tangential lines of sight dedicated to studies of fast ion behaviour, in particular of the redistributionof fast ions in kinetic and real spaces arising from the presence of MHD instabilities, such as
fishbones, sawteeth, NTMs and Alfvén Eigenmodes.
In ITER, in addition to heating, NBI will be used for current drive (NBCD). Fast NBI ions can drive
Toroidal Alfvén Eigenmodes (TAE) leading to their redistribution. These processes affect the
profile of the NBCD leading to a deviation from that predicted in their absence. A multi-lines-ofsight tangential spectrometer that can study the fast ion distribution function will measure thischange in fast ion distribution and associated NBCD profile. Tangential measurements will beinterpreted in combination with data from other existing radial diagnostics in order to reconstruct
the 3-dimensional distribution function. The TNS, based on diamond detectors, could measure fast
charge-exchange atoms directly (if operational in PFPO-2) and will provide collimated spectra of
DT neutrons in the nuclear phase of ITER.
The measurement timescales of the proposed diamond-based neutron spectrometer would allow thedetermination of the fast ion distribution function in the presence of MHD events such as sawteeth,
fishbones, NTMs and TAE modes. The TNS observes ITER’s plasma with three chords that cross i)
plasma core, ii) ¼ of minor radius and iii) ½ of minor radius. Three detector modules, consisting ofdifferent sensitivity diamond detectors will be installed in the central drawer of equatorial port 8.

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Collective Thomson Scattering (55.C7)

The CTS is implemented in Equatorial port 12. Currently, the front-end of this system is being
designed as part of an EU-DA package and is progressing well. The back-end interfaces are being
enabled and finalization of the design will be carried out once the funds become available.
Table H-4 – Diagnostic systems available for FPO (in addition to those for First Plasma,
PFPO-1 and PFPO-2)
PBS
55.B1
55.B2
55.B7
55.B9
55.BB
55.BD

System Name
Radial Neutron Camera
Vertical Neutron Camera
Radial Gamma Ray Spectrometer
Lost Alpha Monitor
High Resolution Neutron Spectrometer
Vertical Gamma Ray Spectrometer

55.BE

Tangential Neutron Spectrometer

55.BV
55.C7
55.GC

Neutron Calibration (14 MeV)
Collective Thomson Scattering
Tritium Monitor

Comment

Captive parts installed
Captive parts installed
Captive parts installed
Captive parts installed
Design Being
Developed
Captive parts installed

H.6 Impact of 1.8 T operation on diagnostic systems and specific upgrades required
The ITER Research Plan considers operation at 1.8 T to access the H-mode in PFPO-1 (section
2.5.4.9) and to further explore it in PFPO-2 (section 2.5.5.7). This has an impact on the ability ofthe diagnostic systems to meet their measurement requirements. Corrective measures can be taken
and the associated cost impact is being evaluated. It is also noted that the ECRH system may
operate at a lower frequency (between 104 and 110 GHz) besides the nominal frequency(170 GHz). Therefore, diagnostic systems will have to be protected from ECRH stray radiation at
both frequencies if this ECRH option is adopted. In this section we highlight the main diagnostic
systems which are impacted by this operation and describe the upgrades that would be required. In
general, with the exception of some microwave systems, there is no plan to modify the hardware
requirements for the diagnostic systems. This implies that, in general, a compromise has to bereached between measurement accuracy and time resolution for some of the plasma parameters to
be measured during 1.8 T operation.
H.6.1 Model plasmas used for the evaluations
To evaluate the impact of 1.8 T operation on the diagnostic systems, the following plasma
conditions have been considered

5 MA/1.8 T H-mode He scenario with 20 MW of ECRH: [ITER_D_TP3QQA, 2016] andassociated DINA output [ITER_D_TMT985, 2016];



METIS and ASTRA simulations for helium H-modes and hydrogen L-mode plasmas (seeAppedix F).

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During the flat-top period, the electron densities are in the range of (1-2)×1019 m-3 while theelectron temperature can have peak values higher than 10 keV in these plasmas. Work is on-goingto evaluate the performance of the diagnostic systems and to determine the required upgrades. The
status of this work is summarized below.
H.6.2 Implications of 1.8 T operation on diagnostic systems
H.6.2.1 ECE (55.F1)
The low frequency radiometer can be used for measurement of the plasma temperature, providing
high quality measurements in the pedestal region. However, an upgrade to the back-end is required
to provide measurements of the 2nd harmonic X-mode, as detailed in [ITER_D_TXQ5E8, 2016].
Note that this diagnostic is available from PFPO-1.
H.6.2.2 Core Thomson scattering (55.C1)
The Core Thomson Scattering system is available in PFPO-1. The minimum density to be measured
by this diagnostic according to SRD-55 [ITER_D_28B39L, 2017] is 31019 m-3 and the baseline
temporal resolution is 10 ms (100 Hz) and for PFPO-1, one laser (between 20 and 50 Hz) iscurrently expected to be delivered. For lower plasma densities than the baseline, such as those for
5 MA/1.8 T operation, measurements are possible by integrating the data over several laser pulses.
Assuming that the system is delivered with a 20 Hz laser, then for a reduced density of 11019 m-3,
the integration would be approximately 3-4 times longer (classically 3, but 4 to account for detector
noise). This would mean that the signal would be provided on a 200 ms basis rather than with
50 ms. If the 50 Hz laser option is deployed in PFPO-1, then the density measurement at 11019 m-3
would have a correspondingly higher time resolution.
H.6.2.3 Edge Thomson scattering (55.C2)
The edge Thomson scattering (ETS) is available in PFPO-2 although there is a plan to implement it
and commission it earlier as described above. If this is possible, then the basic edge electron profile
measurements can be available towards the end of PFPO-1. According to the current baseline, theETS is capable of measuring the electron density within 51018 - 31020 m-3. The temporal
resolution is 10 ms (100 Hz) assuming the full laser capacity is installed. For lower density
measurement, relaxation of the temporal resolution for signal integration is needed as for the core
TS. For instance, the temporal resolution would be > 50 ms to measure electron density close to11018 m-3. Any further low density measurement would be impractical because of thermal noise in
the electronics.
H.6.2.4 LFS-Reflectometer (edge density profiles) (55.F2)
In the current configuration, most of the pedestal would be missed for 1.8 T operation. A mitigation
measure would be to install a new back-end in the diagnostic building covering the Q-band (3350 GHz). The frequency multiplexer should also be adapted. Transmission lines can work withsome degraded performance. Note that this diagnostic would only be available from PFPO-2.
H.6.2.5 Plasma Position Reflectometer (gap control) (55.F3)
For this system, there is no significant impact. This system uses O-mode in its baseline and does notrely on the magnetic field. Low density means that its performance might be slightly degraded.
Note that this diagnostic would only be available from PFPO-2.
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H.6.2.6. HFS Reflectometer (core profile) (55.F9)
To keep a baseline radial coverage, the frequency capability should be lowered by adding a
Ku-band 12-18 GHz. A new back-end frequency multiplexer would be required. The transmission
lines and vacuum windows are likely to limit the performance of the system; this needs further
detailed study. Note that this diagnostic would only be available from PFPO-2.
H.6.2.7 CIXS (X-ray core spectrometer to measure the core Ion Temperature) (55.E5)
Tungsten emission in the 1.8 T discharges would potentially be challenging to measure due to thepotentially low W concentrations. This could jeopardize the ion temperature measurement in
PFPO-1, which are key because the expected high Te/Ti ratios in these plasmas. Note that the otherdiagnostic for ion temperature measurements (CXRS) is not yet available in PFPO-1. As amitigation, the premixing of 0.001% Xe with the fuelling gas is proposed and has been requested to
be included in the gas introduction system (leading to nXe/ne = 10-5 in the plasma), which, according
to modelling, should allow the measurement requirements to be met. Note that this diagnostic isavailable from PFPO-1.
H.6.2.8 XRCS (X-ray Crystal Spectroscopy Edge- to Measure the Edge Ion Temperature)(55.EI)The 55.EI XRCS spectrometer is one of four XRCS subsystems that together measure the ITER
core and edge plasmas. Its primary goal is to measure ion temperature and plasma rotation profiles
in the outer plasma. Poloidal profiles of the plasma at r/a > 0.9 are imaged via a spherical crystal
located outside of an upper port, onto a 2-D X-ray detector. It is optimized for very high spectral
resolution in a narrow band around wavelengths 0.2 - 0.5 nm interval, chosen to containrepresentative line emission from the main impurities. This system is important for the
PFPO-1 phase, as there will be no other edge ion temperature measurement. The impact of 1.8 T in
this diagnostic measurement would be through the level of impurities in the 5 MA/1.8 T H-mode
plasmas.
H.6.2.9 TIP Interferometer (55.C5)
No major signal or accuracy degradation is expected for the interferometry measurement.
Polarimetry measurement is likely not to be available at these densities. Interferometry fringe jumps
would therefore need to be corrected without the help of polarimetry. This is not expected to be amajor issue as the densities are low. Note that this diagnostic will be available from PFPO-1.
H.6.2.10 DIP Density Interferometer/Polarimeter (55.FA)
No major signal or accuracy degradation is expected for interferometry measurement. Polarimetry
measurement is likely to be not available at these densities. This is not expected to be a major issue
as the densities are low. Note that this diagnostic will be available from PFPO-1.
H.6.2.11 MSE Measurement (55.EB)
No impact on hardware and minimal impact on performance are expected. The Doppler shiftchanges as well as the Stark splitting with 5 MA/ 1.8 T but changes are smaller than the changes
expected when going from H-beams (PFPO-2) to D-beams (FPO) and are therefore covered in thedesign. Note that this diagnostic would only be available from PFPO-2.

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H.6.3 ECRH stray radiation protection
All microwave diagnostics have been designed to be protected from 170 GHz stray radiation (and60 GHz from CTS). They need to be protected from ~ 105 GHz radiation if ECRH heating at thisfrequency is to be considered for 1.8 T operation. These protection features will make use of: pinswitches, notch filters, band-stop filters and stray radiation detectors with active feedback to the
ECRH heating:

55.F1-ECE: New filters have to be developed and integrated in the system.



55.F2-LFS-Reflectometer: All the strategy for stray radiation protection has to be changed
to include the 105 GHz range. Back-end protection is mostly done in the diagnostic building
but vacuum windows are mainly protected by the customized grated mitre bend in the port
plug. It is unlikely that it will be possible to change these components for 1.8 T operations.
A strategy should be agreed to avoid 105 GHz ECRH stray radiation arriving around Eq.
Port#11.



55.F3-Plasma Position Reflectometer: New filters have to be developed and integrated in
the system.



55.F9-HFS Reflectometer: New filters have to be developed and integrated in the system.



55.NW-Vacuum windows: In a generic approach, it is planned to protect vacuum windows
by coating the inner surfaces of the in-port plug optical paths at risk of stray ECRH radiation
with a microwave absorber layer. This coating is optimized for 170 GHz absorption. Its
efficiency at 105 GHz should be assessed.

H.7. Conclusions
An explanation of the measurement systems available for each phase of operation in the Staged
Approach with an emphasis on the missing components (upgrades to improve baseline diagnostics
and to enable diagnostics) has been provided in this Appendix. This shows that a basic diagnostic
set is available for First Plasma. Subsequent plasma phases will add diagnostics in line with what isrequired for the development of the ITER Research Plan. The systems that need further
improvement (upgrades) or development (enabled) have been described along with the items thatare missing. In the current planning, the budget for these systems is expected to be available for theoperational phase. Some budgets for critical system development have already been provided toallow interfaces and other critical aspects to be prepared, in advance, for these diagnostic systems.
In some cases, the planning is reasonable as many of the missing items are back-ends and can bedelivered quickly. In most cases, a delay in availability of the funding to provide the upgrade orbuild the enabled diagnostic will cause an increased risk for the systems’ availability with therequired performance (e.g. lower performance for the upgrades or no measurement capability forthe enabled systems).
The second aspect discussed in this Appendix is the ability to measure parameters in 1.8 T plasma
operation and with, possibly, some level of ECRH heating at lower frequency (104-110 GHz). In
this case, the main impacted systems have been studied to give an indication of how they canaddress the 1.8 T operation. The general conclusion is that with small adjustments to some systems
and, in some cases, increasing the time resolution of the measurements, the 1.8 T plasmas will bereasonably well diagnosed. Note that the Edge Thomson scattering is currently planned for PFPO-2
but it is expected to be installed in PFPO-1 and commissioned during this phase. Currently, thereflectometers which can measure edge density profiles are expected to be installed in PFPO-2, but
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these could also be accelerated to make measurements if it is deemed appropriate. A decision on
this could be made on the timescale of the PCR that will implement the 1.8 T operation, which ispresently in preparation.

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Appendix I: Tritium Availability
ITER will procure tritium for the DT experimental program from external civilian sources. The total
mass of tritium which will be consumed to achieve the mission goal of producing an average
neutron wall loading of 0.3 MW.yr.m-2 laid down in the Project Specification [ITER_D_2DY7NG,
2010] is ~ 14.5 kg. In addition, a site inventory of several kg must be maintained for the efficient
operation of the Fuel Cycle systems, and to account for tritium retained in, e.g., plasma-facing
components. The ITER project will not start using tritium on-site until the tritium commissioning of
the Tritium Plant commences (currently foreseen in 2033) and the ITER experimental program will
start to consume significant amounts of tritium from 2036 onwards. To achieve the neutron wallloading target on a timescale commensurate with the lifetime of the ITER Agreement implies a burnrate of approximately 1 kg.yr-1 for ~ 15 years following 2036. The question arises, therefore, as towhether the global civilian tritium inventory during the period 2036 – 2050 will be sufficient tomeet the needs of the project.
Most of the tritium within the civilian stockpile is produced by fission reactors using heavy water asa moderator, commonly referred to as CANDU reactors. Such reactors are in operation for power
production in Canada, Korea, Romania, India, China and Argentina. However, only Canada andKorea are known to extract significant tritium from their reactors which could be exploited‘commercially’, and Romania has plans to do so. Estimates are available for the present stockpile
and the current rate of tritium production in the Canadian and Korean fission programs. However,estimating the long-term tritium inventory from these sources involves several uncertainties, sinceat least some of these reactors will be decommissioned sometime in the next decade and Canada,
which is currently the world’s major producer of ‘commercial’ tritium, appears to have no plans tobuild further CANDU reactors (though it is proposed to refurbish at least some of the CANDU
reactors to extend their operating life). Future estimates of the global tritium inventory, therefore,
indicate that a maximum inventory is likely to be reached in the late 2020s (see Figure I-1 and FigureI-2), with the inventory declining thereafter as the loss of tritium by radioactive decay (tritium has a
half-life of 12.3 years) exceeds the production rate.
Based on internal estimates of the current tritium inventory and production rates in Canada andKorea, as well as the expected lifetime of the CANDU-type reactors, an estimate of the evolution of
the civilian tritium inventory over the next 35 years has been developed with and without ITER
consumption. The results for 2 different assumptions concerning the ITER tritium consumption rate
are shown in Figures Figure I-1 and Figure I-2. The profile of tritium consumption used in thesecalculations is based on the report [Loarte, 2016-1], which was developed to provide an estimate of
the profile of fusion power production over an extended ITER DT program satisfying the neutron
wall loading mission goal. The analysis was carried out for the prevailing assumptions on theexperimental program, in particular that ITER operations would be organized on a 3-shift basis witha 25% burn duty cycle. This led to an estimate of annual tritium consumption (averaged over a 2year operational campaign with 8 months of scheduled shutdown/maintenance) of ~ 1.2 kg. Figure
I-1 shows the evolution of the estimated global tritium inventory when this consumption rate isincluded, starting in 2038 (it is assumed that during 2036 – 2037 DT campaign a nominal amount of~100 g of tritium will be consumed). These figures include the need to build sufficient ITER tritium
inventory for proper operation of the Fuel Cycle systems and the return of this tritium to the world
market at the conclusion of DT operations.
Due to guidance received from the ITER Members on the expected budget for the Operation Phase,
it has been proposed, for the present, to plan ITER DT operations on a 2-shift pattern. This reduces

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the annual neutron production by approximately 25% and, in consequence, reduces the annual
tritium consumption by a similar amount, to ~ 0.9 kg.yr-1. To achieve the lifetime wall loading
therefore requires an extension of the DT campaign to ~ 18 years (including the DT campaign of2036 – 2037), as opposed to 14 years in the case of 3-shift operation. The estimated evolution of thetritium inventory under these assumptions is illustrated in Figure I-2.
Overall, the conclusion from the analysis shown in Figures Figure I-1 and Figure I-2 is that thepresent estimates of global tritium inventory are (just) sufficient for the ITER DT program to meetthe neutron wall loading mission goal, if it assumed that there is no other major consumer of tritium
during the next 35 years. A somewhat more extensive survey and analysis of the global tritium
production and future inventory in the context of the tritium requirements to start-up a Demoreactor was reported recently by [Kovari, 2018]. The results presented indicate a somewhat higher
maximum inventory being reached in the 2020’s (the maximum Canadian inventory is estimated tocorrespond roughly to the ‘global’ maximum shown in Figures Figure I-1 and Figure I-2). Thissuggests that the analysis presented here may err on the conservative side, which is positive fromthe perspective of ITER’s future tritium requirements.

Figure I-1: Predicted evolution of ‘global’ tritium inventory assuming an ITER DT program achieving an
average neutron wall loading of 0.3 MW.yr.m-2 over ~12 years in 3-shift operation.

Figure I-2: Predicted evolution of ‘global’ tritium inventory assuming an ITER DT program achieving an
average neutron wall loading of 0.3 MW.yr.m-2 over ~16 years in 2-shift operation.

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Appendix J: Research Plan Risk Register
The RAMI analysis provides an estimate of the overall availability of the ITER tokamak and its ancillary systems for experimental studies, andestablishes a framework within which a time schedule for the development of the experimental program can be assembled. This time schedule ispredicated on several assumptions about the efficiency of the experimental program, i.e. its ability to meet the key milestones in the progress towards
the Q ≥ 10 goal, and about fusion plasma physics at the ITER scale: essentially that the physics of plasmas at the ITER scale follows the predictions
formulated using the physics basis derived from current tokamak experiments. The rate of progress actually maintained in the experimental program
and the level of fusion performance which can be achieved are therefore subject to a variety of uncertainties, and these give rise to risks of delays in
the program or of failure to meet the principal mission goals within the timescale desired (or even of failure to meet them at all).
A Risk Assessment has therefore been assembled which identifies the top risks to the scientific program associated with possible limitations inexperimental techniques or deviations in plasma behaviour from that predicted. These risks, the potential consequences for the operations timeschedule towards DD/DT operation (targeted for late 2035 within the Staged Approach) and for the achievable fusion performance in terms of thelong-pulse Q ≥ 10 mission, together with the possible mitigation measures which could be implemented, are summarized in the appended table. The
focus in this analysis has been on factors related to experimental techniques and to plasma physics, and no attempt has been made to assess theimplications of limitations in the installed hardware, which is dealt with in the RAMI analysis and in the Project Risk and Opportunity Register.

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1

2

Risk

Research Plan Implications

Mitigation

Inadequate Disruption Mitigation:(i) DMS has inadequate performance, in mitigatingheat loads or runaway electrons;(ii) consequences of DMS use excessive (e.g. firstwall melting, long down time);(iii) inadequate prediction capabilities.

Several potential consequences for Research Plan:(i) limitations in maximum current and fusion power;(ii) more frequent shutdowns to replace in-vessel
components;(iii) excessive number of uncontrolled disruptions
leading to loss of operational time and slower
progress in scientific program.

Focussed R&D on DMS-related experiments inpresent devices and development of validated models
is essential to allow appropriate extrapolation ofrequirements to ITER. This can also contribute todevelopment of improved disruption predictive
capability. Possible mitigating measures related to
ITER operation:

Potential consequences:

Further R&D on H-mode power threshold in current
program and its link to confinement quality is an
essential prerequisite for ensuring that ITER can takethe necessary measures in advance of operation. If
uncertainties remain large, possible mitigation
measures are:(a.i) prepare increase in baseline H&CD capability inadvance of or during operation (latter implies delays);(a.ii) accept risk of proceeding to D operation
without H-mode experience and develop H-mode
under these conditions – potential delay due to morecautious rate of progress and/or need to replacedamaged PFCs.
(b.i) prepare for DT operation in D plasmas at low
current and field – implies increased risk in DT phase(b.ii) increase H&CD power – possibly implies

H-mode power threshold
uncertainty range:(a) in H/He phase(b) in D phase

at

high

end

of

(a) inability to achieve H-mode in H/He plasmas(b) access to H-mode only in restricted range at low
values of plasma current/ field404(i) install flexible DMS (incorporating more than oneof present options) to provide greater ability to tune
system to ITER plasmas and to respond to present
uncertainties;(ii) more cautious approach to scenario development,
likely to lead to delays in program;(iii) additional time allocated to R&D on DMS in
ITER program and implementation of upgrades,
leading to delays in program;(iv) limitations on plasma current and fusion power,with possible consequences for mission goals.

ITR-18-003

3(c) in DT phase(c) baseline H&CD allows access to H-mode overlimited range of current and field

Inadequate ELM mitigation schemes:(i) RMP coils do not have planned mitigation effect,
or operate over restricted range in q95, or cause
unacceptable braking of plasma (lowers locked mode
threshold);

ITER is equipped with at 27 ELM control coils but
the uncertainties in extrapolation to ITER remain
large and this risk must be addressed:(i) complete failure of ELM control coils to mitigate
type-I ELMs (or achieve mitigation without
unacceptable consequences) would necessitate
upgrade of the pellet system for pellet pacing ordevelopment of an alternative operation scenario or
R&D in ITER itself to explore alternative methods of
ELM mitigation;(ii) operation over a restricted range in q95 would
necessitate a focus on the fixed-q95 approach to thedevelopment of the 15MA scenario; this issue raises
potentially serious issues for the hybrid and steadystate scenarios, which operate at higher q95 and might
imply the need to develop operation in an alternative
ELM scenario;(ii) vertical control system not capable of triggering
ELMs, even at low plasma current

405

delays unless decision taken before ITER operation.
(c.i) increase H&CD power and accept possible
limitation of Q < 8 with shorter pulse length due tohigher power operation; higher power operation
could also be more demanding for detached divertor
operation.
R&D involving multiple devices and several
alternative ELM mitigation techniques can reduce theuncertainties in extrapolating to ITER, particularly if
successful models of the mitigation processes aredeveloped. Research should continue to identify andunderstand alternatives to the ‘established’
techniques. Possible mitigating measures related to
ITER operation and their implications include:(i) upgrade of the pellet system to provide ELM
control capability with acceptable fuelling/fuelthroughput;(ii) development of an alternative operating scenario,
which would imply delays in the program as time
would have to be invested in exploring alternative
scenarios in ITER and might ultimately imply
operation at reduced Q;(ii) Development of alternative ELM mitigation
methods within the ITER program, which would
imply delays due to program time required, andpossibly due to the need to install upgrades;(iii) limitations in the applicable range in q95, Bt or ne,
would have consequences for the other main
operating scenarios and might lead to delays in the
program as alternative approaches were explored.

ITR-18-003

4

Inadequate vertical stability control:(i) unacceptable level of noise in diagnostics for VS;(ii) failure of in-vessel coils;

Limitations in the vertical stabilization capability
could result in program delays, since plasma
elongation would likely be limited, restricting thedevelopment of the baseline scenario until solutions
were identified and implemented:(i) an unacceptable level of noise in the VS
diagnostics, where there are significant uncertainties,
would likely limit the usable elongation and the pulse
length due to overheating of in-vessel coils and AC
losses in superconducting coils;(ii) failure of the in-vessel coils would necessitate theimplementation of the foreseen backup option (VS1 +VS2) and an alternative approach to the 15MA
scenario development; the former, in particular,
would result in delays to the program.

406(i) Further R&D to characterize the level of noise in
VS systems in existing devices and to identify theprincipal sources will allow ITER to plan in advancemeasures to limit the level of noise in the VS system.
If such measures are inadequate, the consequence islikely to be a shorter allowable pulse length and, in
extremis, restrictions on the plasma elongation andtherefore current and Q. The program might also
suffer delays as a result of less robust operation.
(ii) implementation of the VS1 + VS2 backup option,
while delaying the program, would, in principle,
allow the baseline 15MA scenario to be maintained.
R&D to date has given confidence that techniques forthe control of the plasma inductance during the pulse
can be used to limit excursions in the stability
margin; nevertheless further research, particularly in
relation to the controlled exit from the H-mode andcurrent ramp-down is desirable to strengthen
confidence on the capability for achieving therequired fusion performance and pulse length. In
addition, development of high current operation in Lmode might be more constrained, requiring a moremeasured development program during the nonactive phase.

ITR-18-003
5Lack of reliable high power heating during nonactive phase of program:

Poor performance from one or more of the heating
systems is more likely to occur early in the program,i.e. during the non-active phase, and would havemore severe consequences; the most severe
consequence would be an inability to achieve an Hmode under any conditions, implying both (i) theneed for the immediate implementation of a H&CD
upgrade program and (ii) either delay in the nonactive phase program, or deferral of the H-mode
program to the deuterium phase.

407

The extensive R&D program foreseen for ITER
H&CD systems and the various design changes
incorporated during the Design Review are alreadysignificant mitigation measures; nevertheless,progress in the research program remains heavily
dependent on reliable high power operation of the
H&CD systems; further mitigation measures in theevent that reliability is an issue early in the program
would include:(i) an immediate upgrade program for one or more
H&CD systems, which would nevertheless require
several years and likely delay the transition to DT
operation;(ii) accept risk of proceeding to D operation without
H-mode experience and develop H-mode under these
conditions – potential delay due to limitations in nonactive operation and need to perform required
program after start of nuclear operation.
(iii) prepare for DT operation in D plasmas at low
current and field – implies increased risk in DT
phase.
A combination of (ii) and (iii) in parallel with theimplementation of (i) might allow a higher risk pathto DT operation with minimal delay relative to the
proposed schedule.

ITR-18-003
6Unacceptable ‘divertor’tungsten PFCs:

Any failure to reach the level of divertor performancerequired can delay the program as alternative regimes
are sought and might necessitate more frequent
divertor PFC replacement:

An expanding program of R&D on the use of high-Z
PFCs is being implemented in both the major fusion
devices and in laboratory facilities which will greatly
improve the understanding of tungsten a high heat
flux PFC and of the transport and possible
accumulation of tungsten in the divertor, SOL andcore plasmas. It is essential that the level and scopeof the ongoing R&D be maintained. Nevertheless,given the uncertainties in, e.g., transport physics andthe scale of the extrapolation to ITER, e.g., intransient thermal loads, it is inevitable that the use oftungsten for high heat flux components will have anexperimental aspect, involving issues which will need
to be addressed during ITER operation(i) limitation of the capability to establish a radiating
divertor with acceptable core impurity contamination
will likely limit the Q which can be achieved andpossibly shorten the maximum burn if accumulation
occurs throughout the pulse;(i) high power core electron heating, which
dominates in ITER might be adequate to limit the
impurity concentration and prevent accumulation; if
not, and if the problem is limited to the level ofseeded impurities required for operation withtungsten, this would result in program delays as
techniques are developed to limit core contamination.

(ii) tungsten concentration in the plasma core isunacceptably high, even with high power coreelectron heating;(ii) a high, but stable, tungsten concentration would
limit the Q and shorten the burn duration;
accumulation would likely cause termination of thepulse, making the scenario untenable as a route tohigh-Q operation;(ii) research program would have to be implemented
within the Research Plan to optimize core anddivertor plasma scenarios to minimize tungsten
contamination of core plasma – additional research
activities would delay program to achievement ofhigh fusion power and high-Q;(iii) mitigation of thermal transients is inadequate toprevent enhanced erosion rates and reduced lifetime;(iii) inadequate mitigation of thermal transients will
inevitably delay the program until a satisfactory
solution is found, or delays and costs will accumulate
due to more frequent replacement of the divertor;(iii) additional program time might be required to
improve mitigation techniques for thermal transients;
more frequent replacement of the divertor would benecessary, implying delays and additional cost;(iv) reliable plasma start-up cannot be achieved due
to high-sputtering rates in initial low density plasma;(iv) failure to achieve reliable plasma start-up would
either delay the program, while a more limited, but
acceptable, window for the start-up was developed;
this could have implications for control and for thedevelopment of more advanced scenarios if it limited
the scope for heating and current drive during the(iv) R&D in existing devices can contribute topreparations for a suitable start-up scenario, but
dedicated experimental work will be required in
ITER to develop a regime which is appropriate to the
ITER plasma parameters; success would lead to a
rapid development of suitable scenarios to meet theperformancewith(i) impurity seeding to produce radiative divertor
leads to excessive core impurity contamination;

408

ITR-18-003

7current ramp-up;

ITER mission, but if difficulties with impurity
generation are encountered, it will cause delays in the
program and possibly limit the flexibility ofoperation; subsequent development of hybrid andsteady-state scenarios would also be affected;(v) partial melting of W-targets limits power handlingcapability and prevents satisfactory high power
operation.

(v) if partial melting of the tungsten target cannot be
avoided, it is likely to lead to a delay in the program
either to replace areas of the divertor PFCs (if these
are infrequent events), or to develop improvements inoperational strategies.

(v) the cause of the localized melting would influence
the possible mitigation measures: inadequate
mitigation of thermal loads would imply either
improvements in the mitigation techniques, or if thiswere unsuccessful (see Risks 1 and 3) more frequent
replacement of the affected areas; if the melting werecaused by failure of the divertor detachment control,
additional program time would be required todevelop more robust control of the divertor plasma
parameters.

(vi) operation with helium plasmas could lead toenhanced erosion or to small scale damage totungsten PFC surface, leading to shortening of PFC
lifetime and possible limitation of stationary powerhandling.

(vi) this could lead to need, e.g., for extensive
replacement of tungsten PFCs before nuclear phase
operations (implying additional cost) and/or
limitations on divertor PFC power handlingcapability; alternatively, a decision not to carry out
experiments with helium would mean that H-mode
and ELM control experiments may not be attempted
until deuterium phase, implying a probable delay inexperimental program.

(vi) Further R&D in present program will be required
to quantify how serious such effects will be and theimplications for ITER operation; if substantial helium
impact on surface morphology of tungsten PFCs isconfirmed, leading to significantly increased erosion
rate, and if hydrogen H-modes are not possible in
PFPO, the delay of H-mode experiments todeuterium phase would appear to be only solution atpresent; overall delay in experimental program would
then have to be analyzed.

(vii) Recrystallization of tungsten in the high heat
flux area of the divertor increases the risk of macrocrack formation(yii) this would result in a lowering of the allowable
heat fluxes in the divertor and possibly a restriction in
the ITER high-Q operational domain (in terms of
power and pulse length) or the need to operate with amore deeply detached divertor most likely increasing
the disruption risks.

(vii) an operational budget will be defined for thedivertor to account for the fact that tungsten
recrystallization is a time and temperature-dependent
effect. This will define how long the divertor can be
operated at a given heat flux density and ensure a
tracking of its integrity. R&D is planned to study indetails the recrystallization kinetics of the material to
be used for the ITER divertor.

Level of toroidal field ripple degrades plasma
performance:

Potential consequences:409

Significant R&D in this area is in progress, though aquantitative basis for extrapolation to prediction ofconfinement effects in ITER has not yet been
established; DIII-D experiments with a TBM mock-ITR-18-003
up have shed additional light on these effects and therole of localized field perturbation and allow an
improved specification for the TBM configuration soas to reduce any deleterious effects; a key issue in
ITER would be to show unambiguously that any
shortfall in confinement is due to ripple effects;(i) residual TF ripple after installation of ferritic
inserts (~0.6% in non-regular sectors) leads todegradation of energy and particle confinement;(i) reduction of energy confinement would imply arisk that Q=10 could not be achieved in 15MA
reference scenario;(ii) localized ripple due to TBMs either degrades
plasma confinement or causes plasma braking and/or
error field growth.

(ii) localized ripple might degrade confinement as in(i), while plasma braking and localized error fields
could give rise to locked modes reducing reliabilityof operation.

(i) if TF ripple is beyond the level at whichsignificant confinement effects are manifested, little
can be done in mitigation without removing theblanket modules to attach ferritic material to thevacuum vessel wall, at considerable delay to theprogram; a more productive approach would be toseek an optimized equilibrium which minimizes theinfluence of TF ripple on confinement; alternatively,
options for scenarios with enhanced confinement
could be explored; both approaches would require
additional program time;(ii) the availability of dummy port plugs and nonferromagnetic TBMs would allow the TBMs to bewithdrawn for some period to permit the program
towards DT operation to proceed; apart from any lossof time in identifying the deleterious effects and in
replacing the TBMs, the impact on the physics
program could be limited; however, this would
constitute a significant setback for the TBM Program,with potentially long delays unless alternative
mitigation techniques were identified;(iii) At least a part of the error field and plasma
braking effects due to TBMs can be corrected using
the error field correction coils (and possibly, ELM
control coils).

8

Lack of plasma rotation leads to degradation ofplasma performance:

Potential consequences:

410

R&D is ongoing in present experiments to improveunderstanding of intrinsic rotation and momentum
transport in tokamak plasmas, but it is a longstanding issue in fusion research; there is a strong

ITR-18-003
physics basis for the influence of error fields androtation is clearly an important aspect of thisproblem, but the quantitative influence remains
uncertain. Continuing R&D in this area will be
important for improving ITER’s capability to deal
with error fields, but the change in scale which ITER
brings may also reveal new insights or issues.
(i) intrinsic rotation of ITER plasmas is inadequate toachieve confinement quality consistent with H=1;(i) reduction of energy confinement would imply arisk that Q=10 could not be achieved in 15MA
reference scenario;(i) there is limited scope for increasing torque in
ITER (e.g. the installation of the 3rd HNB); if low
rotation is identified as an issue, some mitigation may
be possible by accurate correction of residual error
fields using both internal (ELM control coils) andexternal (correction coil) saddle coil systems – themajor impact on the schedule might be associatedwith the time needed for accurate measurement andcorrection of error fields; if the plasma rotation
remains inadequate after careful correction of error
fields, the installation of a low (~100 keV) positiveion beam would provide a possible solution, but thiswould imply a significant delay to the program toprocure and install such a system before the tokamak
environment became active.

(ii) low intrinsic rotation makes ITER plasmas
particularly sensitive to error fields, either due toassembly tolerances or distribution of magnetic
material, leading to growth of locked modes andhigher than expected disruptivity.

(ii) additional program time might be required toanalyze error field sources and improve level ofcorrection to allow more reliable operation at
q95=3;.most severe consequence could be restriction
in maximum current (lowest q95), limiting fusion
performance.

(ii) ITER is equipped with a significant error fieldcorrection capability via an array of 18 external
saddle coils – the quality of correction might welldepend on the time dedicated to determining themagnitude and distribution of error fields, andadditional time, including the use of dynamic
detection techniques, could be allocated if required;
the internal ELM control coils could provide
additional correction capability, at least on an
experimental basis, to determine whether the externalcorrection set is adequate; any upgrade in this area
would imply a significant delay in the program.
(iii) Error field and plasma braking effects due to
TBMs can be partially corrected using the error fieldcorrection coils (and, possibly, ELM control coils).

411

ITR-18-003
9High levels of tritium retention require more
frequent tritium removal procedures thanforeseen:(i) Research program is slowed by need for frequent
vessel conditioning for tritium recovery.

(i) if Be-codeposition dominates the tritium retention
rate, a longer-term solution would involve replacing
Be first wall surfaces with tungsten, initially inregions not in contact with plasma (i.e. exposed only
to neutrals) to test that plasma performance is notdegraded by the presence of tungsten on the firstwall; increasing areas of the wall could be replaced
over time – at least in the medium-term, until the
Q=10 mission were accomplished, program delays
due to tritium recovery (assuming suitable techniques
have been developed) might be more acceptable thandelays involved in exchanging first wall materials.

(ii) Efficiency of the baseline T-removal strategy(baking of first wall at 240oC and divertor at 350oC)
is too low for efficient tritium recovery on thickberyllium co-deposits(ii) very frequent and/or long bake-outs would benecessary, which is penalizing for the machine duty
cycle(ii) R&D is ongoing to refine the understanding of T
trapping in Be co-deposits. In addition, several
methods are being investigated in the fusion
community for alternative T-removal techniques.
Performing divertor baking experiments early during
PFPO operations will allow checking the fuelremoval efficiency and confirm whether additional
techniques would be required.

(iii) Significant tritiated ammonia formation duringnitrogen seeding in the divertor, contributing to thetritium retention on divertor cryopumps(iii) Ammonia is efficiently trapped on the active
charcoal used in the divertor cryopumps and requires
an elevated regeneration temperature for its release(similar to water-like gases) which is very timeconsuming.

(iii) Significant R&D is ongoing in current fusion
devices and laboratory to understand where ammonia
is formed in the divertor and predict how much
would be formed in ITER.

(i) tritium retention rate in Be/W environment ishigher than anticipated due to tritium retention via
Be-codeposition.

412

ITR-18-003
10Incompatibility of core plasma requirements for
Q=10 with radiative divertor operation and Q = 5
long-pulse/steady-state scenarios:

Since the divertor peak heat load is limited to
10MWm-2 in long-pulse operation due totechnological constraints, fusion power, and probably
Q would have to be reduced, to allow operation at an
acceptable radiation fraction, until a suitable
alternative operating regime is identified; theminimum consequence would be a delay to theprogram. In addition the Q=10 and Q=5 missions
would also be at risk unless an alternative operating
regime is established.

(i) pedestal temperature requirements for acceptablecore energy confinement incompatible with edge
density for radiative divertor operation;(i) A high pedestal temperature and pressure are
essential to the achievement of the long-pulse Q =5
scenarios, therefore this puts at risk these scenarios
most likely preventing the achievement of the steadystate goal.

(ii) excessive core radiation prevents access to H=1
H-mode operation;(ii) H > 1 is essential for the Q = 5 long-pulse andsteady-state scenarios. This would limit Q and/or
pulse length in these scenarios(iii) divertor asymmetries excessive at requiredradiation fraction, preventing stable operation;(iv) degradation of core confinement in reference
ELMy H-mode associated with use of all-metal PFCs

While many aspects of radiative divertor operation
are investigated in existing devices, it is only at the
ITER scale that the particular combination of edge
and core parameters required for high-Q burning
plasma operation can be established and investigated:
it is likely, therefore, that mitigation measures would
need to be developed in ITER:(i) development of an alternative operating scenario
would imply delays in the program as time would
have to be invested in exploring options foralternative scenarios in ITER; this issue mightultimately imply operation at reduced fusion power
and Q;(ii) exploration of possible alternative radiating
impurities might allow an optimum balance tominimize degradation of fusion performance;(iii) any change in, e.g., divertor geometry
contemplated to counteract this issue would imply adelay to the program in order to design, fabricate andexplore operation in a new divertor configuration;
program time would be required to establish any
influence of divertor geometry;(iv) may need to develop alternative plasma scenario,
for Q =10 resulting in delay to program.
(v) The Q = 5 long-pulse goal could be achieved withnormal inductive H-mode plasmas with Ip = 13 15 MA if H ~ 1.

413

ITR-18-003
11Inability to achieve densities near Greenwald
value required for Q = 10 and Q = 5 longpulse/steady-state:

Several potential consequences for Research Plan:(i) Gas fuelling does not allow adequate density to beachieved in helium plasmas to permit adequate
heating e.g. NB shine-through, ICRF coupling
resistance) or acceptable H-mode behaviour;(i) strategy of using helium H-modes to prepare
scenarios and ELM mitigation techniques in advanceof deuterium operation would not be viable;(ii) fuelling throughput does not allow required
densities to be achieved in DT for Q = 10 and Q=5
scenarios;(ii) reduced fusion performance due to lower density
in high-Q scenarios;(iii) existing technology is unable to fuel plasma
adequately in DT for Q = 10 and Q = 5 scenarios.

12

Inadequate particle control to sustain high-Q
plasma scenarios and Q = 5 long-pulse/steadystate scenarios:(iii) reduced fusion performance due to lower density
in high-Q scenario.

Several potential consequences for Research Plan:

414

While more emphasis is required in existing R&D
program on fuelling and particle transport, thesignificant change of scale and parameters in ITER
implies that fuelling behaviour, in particular, may bequalitatively different, with gas fuelling ineffective,at least for hydrogenic species. In this respect ITER
is accessing a new regime. Possible mitigating
measures for this risk in ITER operation:(i) if helium gas fuelling ineffective, more emphasis
on development of scenarios in hydrogen, where
pellet injection available would be required; ifhydrogen H-modes are not viable, accept risk ofproceeding to D operation without H-mode
experience and develop H-mode under these
conditions – potential delay due to time required forrelevant experimental studies to D phase.
(ii) maximum fuelling throughput can be doubled forshorter pulse durations without upgrades, possibly
allowing Q=10 mission to be maintained with shorter
burn duration but not the Q = 5 long-pulse/steadystate goals; expanding the tritium plant to increase
throughput would be a costly and lengthy mitigation
measure;(iii) pursue options for more effective core fuelling –burning plasma physics could be explored at lower Q
and in various scenarios while necessary concepts aredeveloped;
Overall, inefficient fuelling would be a significant
challenge to the development of magnetic fusionenergy.
Current experiments have generally been able tomaintain adequate particle control for the pulse
lengths and plasma conditions which can beaccessed, and therefore this issue is likely to be
associated with the specific parameters of the ITER
plasma, implying that investigation and mitigation
measures will have to be developed in ITER:ITR-18-003(i) divertor detaches at much lower neutral pressure
than expected, degrading core energy confinement
and helium exhaust capability;(i) required core density for Q ~10 and Q ~ 5
operation cannot be sustained;(ii) divertor helium enrichment lower than expected,
requiring higher particle throughput to maintainrequired exhaust rate;(ii) sustainable burn duration in high-Q conditions
might be significantly shortened;(iii) particle recycling properties of PFCs change
substantially at confinement transitions or due to highpower,
long-pulse
operation,
overwhelming
capability of exhaust systems and resulting in loss ofdensity control and reduction of plasma temperatureand current drive efficiency(iii) high-Q burn conditions cannot be maintained
an/or long-pulse capability is lost due to lower
current drive.;

415(i) present expectations are based on modelling using
a limited pumping capability (the pumps are throttled
to reduce pumping speed – a further factor of 2-4 isavailable), so that first mitigation measure would be
simply to take advantage of the full pumping
capability; if this proved inadequate, the number ofcryopumps would have to be increased, implying
upgrades and a reorganization of the divertor port
allocation; this effect should be observed earlyenough in the experimental program that mitigation
measures could be developed in a timely fashion,
minimizing delays to the DT program – the later it isencountered, the more serious the delay andconsequences for the high-Q mission;(ii) as in (i) headroom in pumping capability isavailable before the number of cryopumps would
have to be increased; the possibility of doubling thefuelling throughput in short pulse operation could
also be exploited; delays to the program would beincurred in developing the scenario, but upgrades
may be avoidable;(iii) as in (i) headroom in pumping capability isavailable before the number of cryopumps would
have to be increased; the possibility of doubling thefuelling throughput in short pulse operation could
also be exploited; delays to the program would beincurred in developing the scenario, but once again
upgrades may be avoidable;
Overall, inefficient particle exhaust would be asignificant challenge to development of fusionenergy, but modelling suggests that ITER has someheadroom in this respect before any upgrades to theinstalled capacity are required.

ITR-18-003
13

NBI current drive efficiency lower than expected
due to fast ion losses by interactions with Alfvén
Eigenmodes

Several potential consequences for Research Plan:(i) Cannot produce or sustain j(r) profile required forhigh/full non-inductive operation(ii) Pulse length for long-pulse Q ~5 goal would be
difficult to achieve(iii) Steady-state Q ~5 goal may not be possible

14

Inadequate core MHD control (NTM, RWM)

(i) Improve fast ion diagnostic measurements in
PFPO-2 (when NBCD will be measured) to:- assess magnitude of ion loss, and consequences for
CD, and.
- to develop control/mitigation schemes (e.g. with
ECRH) as soon as possibleii) Q ~ 5, 1000s goal can be achieved by inductive
plasmas Ip = 13 - 15 MA.

(iii) Upgrades of H&CD will need to be reconsidered
to increase NBCD (if schemes in PFPO-2 are
successful) in addition to those for the Q ~ 5 steadystate scenario itself.

Several potential consequences for Research Plan:(i) Impact on driven current and pulse duration for
long-pulse scenarios;(ii) If more ECRH is required for NTM control less
ECRH power for CD in long-pulse scenarios would
be available thus decreasing pulse length andperformance;(iii) If RWM cannot be controlled then N will belimited preventing the achievement of Q ~ 5 in longpulse scenarios.

416(i) Additional CD capabilities would need to beinstalled. This should recover the pulse length but
may decrease the achievable Q under 5;(ii) More ECCD would need to be installed. Thisshould recover the scenario N but may decrease theachievable Q under 5;(iii) Implement/improve RWM control capability
or/and install 3rd HNB to increase input torque andplasma rotation.