IRPworkbook_USBPO_012622_withFacilityComments_clean

	A	B	C	D	E	F	G	H	I	J	K
1	Ref.	System/ Issue	Required R&D	Category*	Required experimental facilities	Comment	Phase when system required/ Most impacted Phase	US contribution	US collab opportunity	USBPO notes	Version updated by OS, 2020/01/25, with update from facility RP by Loarte et al.
2	Ref.	System/ Issue	Required R&D	Category*	Required experimental facilities	Comment	Phase when system required/ Most impacted Phase	US contribution	US collab opportunity	USBPO notes

3	A. R&D for design completion										Marked as research priority choice from IO for next three years
4	A.1	SPI-single injector. Pellet injection optimization for RE avoidance (incl. TQ and CQ mitigation)	Optimization of shard size, velocity, amount, gas vs. shard fraction, composition (D + impurity) to achieve RE avoidance with optimum TQ, CQ (incl. wall loads)	1	Range of tokamaks with different sizes and plasma parameters to allow extrapolation to ITER (including high Ip tokamak) and with appropriate measurement capabilities	More details on R&D work plan for DMS (https://user.iter.org/?uid=WEE89R)	PFPO-1	US1	C1	D3D leads on this. JET doesnt do multiple SPI anytime soon	!
5	A.2	SPI-single injector demonstration for runaway mitigation	Determination of feasibility to dissipate the energy of formed runaway beams (amount, assimilation) and to improve scheme	1	Range of tokamaks with different sizes and plasma parameters to allow extrapolation to ITER and with appropriate measurement capabilities	More details on R&D work plan for DMS (https://user.iter.org/?uid=WEE89R)	PFPO-1	US1	C1	D3D leads on this. JET doesnt do multiple SPI anytime soon	!
6	A.3	SPI-multiple injections	Determination of effectiveness of multiple injections to achieve RE avoidance with optimum TQ, CQ (incl. wall loads) compared to single injections (incl. timing requirements)	1	Range of tokamaks with different sizes and plasma parameters to allow extrapolation to ITER with at least two injectors from the same/similar locations (toroidal separation not required) and with appropriate measurement capabilities	More details on R&D work plan for DMS (https://user.iter.org/?uid=WEE89R)	PFPO-1	US1	C1	D3D leads on this. KSTAR does have multiple SPI since 2020. Jet possibly in extension.	!
7	A.4	SPI-multiple injectors	Determination of effectiveness of multiple injection from different spatial locations to achieve RE avoidance with optimum TQ, CQ (incl. wall loads)	1	Range of tokamaks with different sizes and plasma parameters to allow extrapolation to ITER with at least two injectors toroidally well separated and with appropriate measurement capabilities	More details on R&D work plan for DMS (https://user.iter.org/?uid=WEE89R)	PFPO-1	US1	C1	D3D leads on this. JET doesnt do multiple SPI anytime soon	!
8	A.5	DMS – alternative injections techniques	Demonstration of the feasibility of the technique to inject material in a tokamak and comparison of mitigation efficiency with SPI	1	Single tokamak demonstration and with appropriate measurement capabilities	More details on R&D work plan for DMS (https://user.iter.org/?uid=WEE89R)	PFPO-1	US1-2		Other US ideas exist. Already leading with Shell Pellet on D3D. NSTX could do something new. Ideas (eg CPP input) include Electromagnetic Particle Injection, Marshall gun, Linear Induction Motor injection. Potential opportunity to expand efforts and world leadership.	Ongoing leadership exploring an alternative DMS
9	A.6	DMS – alternative disruption mitigation strategies	Exploration of disruption mitigation by schemes other than massive injection of D2 and high Z impurities	1	Single tokamak demonstration and with appropriate measurement capabilities	More details on R&D work plan for DMS (https://user.iter.org/?uid=WEE89R)	PFPO-2	US2		US ideas exist, see eg Transients workshop and CPP input. These include Runaway electron (RE) control using RF-induced kinetic instabilities, controlling the current quench with 3D-field induced islands and passive 3D coil for RE deconfinement	Ongoing leadership exploring an alternative DMS
10	A.7	Laser Induced Desorption for in-situ T retention measurement	Demonstrate LIDS as quantitative in-situ diagnostic measurement for T retention in Be co-deposits at divertor	1	Demonstration in tokamak with Be/W environment	Required to provide in-situ measurements of T retained in divertor Be co-deposits (most likely after each operational day)	FPO	US5	C2	ITER demonstration needs Be/W for which JET uniquely is capable. But more generally, LIDS quantification requires establishing methodology for crater quantification and correlation to mass spectrometry. US does have capabilities on this technique and modelling; needs more effort. Technique can be developed elsewhere, collaborations on WEST and with FZ Juelich, Germany re W7-X and JET
11	A.8	Single crystal mirror testing	Performance of single crystal mirror with/without active cleaning	1	Demonstration in Be/W environment	Required for evaluation of performance of ITER diagnostics using plasma facing mirrors	PFPO-1	US5	US2	Requested demonstration needs Be PFCs; JET uniquely capable. Again more generally, work is needed and ongoing on the technique. LIBS Quantification is limited by dissimilar laser-induced ablation mechanisms for Be, C and hydrogen isotopes. US has collaborations on WEST and EAST.
12	A.9	Laser Induced Breakdown Spectroscopy	Demonstrate LIBS as quantitative measurement for T retention in Be co-deposits on main wall	1	Proof of principle demonstration in Be tokamak environment	Can provide an in-situ measurement of T retention in the first wall during shutdown by installation in a robotic arm	FPO	US5	C2	needs Be. Collaborations similar to A.7
13	A.10	SPI-single injector. Pellet injection geometry optimization for RE avoidance (incl. TQ and CQ mitigation)	Optimization of injection direction to achieve RE avoidance with optimum TQ, CQ (incl. wall loads)	2	Range of tokamaks with different sizes and plasma parameters to allow extrapolation to ITER and with appropriate measurement capabilities	More details on R&D work plan for DMS (https://user.iter.org/?uid=WEE89R)	PFPO-1	US1	C2	D3D is doing. Collaboration opportunity - AUG is planning multiple directions.	!
14	A.11	Develop capabilities to measure fast ion losses	Demonstration of quantitative measurements in a tokamak environment (ICE) and/or of compatibility with ITER operation (FILD)	2	Demonstration in suitable tokamak with fast particles and under ITER-relevant conditions	Required to provide a direct measurement of fast ion losses in ITER	FPO (in present plans)	US3	C2	ITPA JEX EP 9. AUG doing a lot of this. PPPL collaboration on JET. GA collab on AUG. DIII-D is doing fast ion loss detection and PPPL does loss orbit calculations using SPIRAL, but these are not unique - the ASCOT code is leading the way for both tokamak (AUG) and stellarator loss orbit calculation. DIII-D does detect ICE, but others in the world are doing the same. There is need improve ICE interpretation. Direct alpha-loss diagnostics (FILD, neutrons) limited to non-US facilities. DIII-D efforts to adapt FIDA interpretation to (primarily low-end of) alpha distribution (W. Heidbrik).	whats planned w.r.t. diagnostics here, any information, involvemenet?
15	A.12	Ammonia formation	Determination of ammonia formation during nitrogen seeded plasmas	2	Divertor tokamaks with metallic PFCs	Provides useful input to the fuel processing plant in ITER	FPO	US5		Needs mettalic PFCs
16	A.13	Neutron diagnostics	Demonstration of measurement capabilities for time of flight 14 MeV neutron spectrometer	2	Tokamaks with sufficient 14 MeV production	Provides input to diagnostic design to provide D/T ratio from neutron measurements	FPO	US4		JET is better suited and we learned is doing this.
17	A.14	IR measurement with reflections in metallic environment	Demonstration of reflection-robust IR temperature measurements of plasma facing components	2	Tokamaks with metallic PFCs and suitable IR systems	Provides input to diagnostic design/optimization and data processing to minimize consequences of reflections on PFC surface temperature determination	PFPO-1	US5		Needs metallic PFCs. WEST and AUG doing.
18	A.15	Radiant Tolerant Detectors	Demonstrate of compact long life detectors for x-ray and VUV	2	x-ray sources combined with neutron and gamma ray sources	Extend the operating capability and availability of these systems	PFPO-2	US2	C2	This issue is important for x-ray diagnostics, which are partly a US responsibity. PPPL researchers are working with manufacturers on R&D, and a collaboration is underway with CERN, where high-luminosity 2020-2021 campaigns expect to use detectors withstanding normalized neutron fluences of 10^15 - 10^17 n/cm2
19	A.16	Neutron Detector development	Compact solid state long life radiation tolerant neutron detector development for in port and invessel e.g. advanced self powered neutron detectors	2	neutron laboratory	Extend the operating capability and availability of these systems	PFPO-2	US3		Neutron diagnostics are an IO responsibility. JET may be well placed to address this given its upcoming DT campaign.
20	A.17	Polarimetric Thomson Scattering	Demonstration of a working Polarimetric Thomson Scattering on a high temperature device.	2	High temperature plasma device >10keV	extend the dynamic range in temperature of a classic thomson scattering system	PFPO-2	US3		D3D Te range marginal. JET better suited.
21	A.18	X-ray optics	Develop x-ray reflection systems to allow extended spatial coverage and reduce neutron transmission.	2	X-ray optics laboratory and acces to a facility to test the components	extend the spatial coverage and detector lifetime	PFPO-2	US2	C2	Day 1 system is IN diagnostic, but US has most expertise, has been asked to take on R&D. US responsible for imaging system. US personnel are not currently allowed to work on design which is causing problems and risk.
22	A.19	Two-wavelength Thomson scattering	Demonstration of a working 2-wavelength Thomson Scattering system on a high temperature device.	2	Experience in Thomson scattering as well as appropriate facilities such as high electron temperature device and suitable experts	Extend the dynamic range in temperature of a classic Thomson scattering system and enable an auto-calibration procedure	PFPO-1	did not rank	did not rank	strikes me (OS) as a field of strength in US program, but not dedicated as diag activity to US, who leads?
23	A.20	LaB6 pressure gauges	LaB6 electron emitter development for ITER relevant conditions	1	Vacuum facility with magnetic field	Extend pressure range and total run time. Evaluate radiation hardness	PFPO-1	did not rank	did not rank	German leadership, ASDEX NGMs with LaB6 are foressen (OS)
24	B. Implementation of the ITER Research Plan
25	B.1. Disruption characterization, prediction and avoidance (for mitigation see Section 1)
26	B.1.1	Disruption/VDE thermal load characterization	Characterization of thermal loads during TQ and CQ (magnitude, time dependence and distribution)	2	Range of tokamaks with metallic walls to minimize radiation from carbon during CQ	Determines plasma operational range in which unmitigated disruptions do not cause melting to PFCs and contributes to the determination of an operational disruption budget	PFPO-1 (it is assumed that DMS will be very effective after PFPO-1)	US2-model U4-expt	C2	Active, world-leading modeling efforts. But no metal wall experiments. JET, AUG collaborations	!
27	B.1.2	Disruption/VDE mechanical load (current flow) characterization including toroidal rotation	Characterization of halo currents during disruptions and VDEs (magnitude, time dependence and distribution, including rotation)	2	Range of tokamaks with a range of vessel conductivities to determine influence of vessel/PFC current path versus plasma physics	Determines plasma operational range in which unmitigated disruptions do not cause category II forces in ITER and contributes to the determination of an operational disruption budget	PFPO-1 (it is assumed that DMS will be very effective after PFPO-1)	US2	C2	Active, world-leading modeling efforts. Capability to measure wall currents not great.	!
28	B.1.3	Runaway electron load characterization	Characterization of power deposition to PFCs by runaway plasmas (magnitude, time dependence and distribution, including magnetic to kinetic energy conversion in termination)	2	Range of tokamaks that can produce reliable runaway beams, vary their terminations and measure power fluxes	Determines plasma operational range in which unmitigated runaway beams do not cause PFC melting and contributes to the determination of an operational disruption budget	PFPO-1 (it is assumed that DMS will be very effective after PFPO-1)	US2	C1	Active, world-leading modeling efforts. Very good RE diagnostics now.	!
29	B.1.4	Disruption detection	Development of disruption detection schemes that are portable across tokamaks	2	Range of tokamaks performing systematic experiments to emulate ITER-like disruptions and to demonstrate routine application of detection scheme	Reliable detection schemes are essential for the practical implementation of disruption mitigation (TQ, if possible, if not at least for CQ mitigation and RE avoidance)	PFPO-2/FPO (this is essential to have a reliable DMS strategy that should be routine before high Ip operation is attempted in PFPO-2)	US3		Not sure of unique US capabilities here. We can and should contribute along with other tokamaks.	Why not listed in Loarte report? -> operational items that are neeed, but not scheduled for completion before comissioning, didnt make the bar in terms of immediate priority due to urgency before comissioning
30	B.1.5	Disruption prediction	Development of disruption predictors that are portable across tokamaks and require minimum re-training	2	Range of tokamaks performing systematic experiments to emulate ITER-like disruptions and to demonstrate routine prediction of disruptions	Reliable predictors are essential for the practical implementation of disruption mitigation	PFPO-2/FPO (this is essential to have a reliable DMS strategy that should be routine before high Ip operation is attempted in PFPO-2)	US2	C1	US3. Lots of ongoing activity, not sure of unique demonstrated capability yet though. Sabbagh others are already doing this and exporting. Not in principle unique but everyone else is doing models with less physics.	dto
31	B.1.6	Disruption avoidance	Development of active operational schemes to avoid disruptions in ITER	2	Range of tokamaks performing systematic experiments to emulate ITER-like plasmas with ITER-like actuators	This involves schemes such as: - to recover plasma thermal stability in high radiative fraction conditions - Pre-emptive application of localized H&CD to prevent growth of MHD that eventually trigger disruptions - etc.	FPO and PFPO-2 (this is most important for high current/power operation near operational/physics limits)	US2	C2	Lots of work here on DIII-D, some unique. Some unique control modeling also. Some ITER-like actuators in D3D. Mulitple collaboration opportunities.	dto
32	B.2. Stationary H-mode plasmas, ELMs, ELM control and impact on H-mode and power fluxes
33	B.2.1	He H-mode operation with W PFCs and ELM control	Compare He H-mode plasmas with W PFCs pedestal behaviour, ELM control and W operational issues (W accumulation, PMI issues) with D H-modes. Highest priority is with 3-D fields but other ELM control schemes such as hydrogen pellet pacing and vertical movements are also important	1/2	Tokamaks with W divertor PFCs capable to operate with He H-modes and to investigate ELM control	Required to determine how to relate H-mode operational experience including W control and ELM control in He plasmas to D/DT. If experiments show that this relation is difficult to establish or that risks due to PMI are too high (see below), including He H-modes in PFPO may have to be reconsidered	PFPO-2 (also affects PFPO-1 but largest impact is on PFPO-2 in which available heating is expected to provide much wider operational space for He H-modes)	US3	C2	Do not have W divertor, but do have ELM control. C wall reference. Data have been taken, could compare w AUG
34	B.2.2	Mixed H + ~ 10 % He H-mode operation and ELM control	Establish H H-mode plasmas with a range of He concentrations to determine the requirements for H-mode access and sustainment and for ELM control schemes compared to H plasmas	1	Tokamaks of various sizes and PFCs and H&CD mixes capable to operate with H + 10 % He H-modes and to investigate ELM control	Required to determine if presence of ~ 10% He can widen H H-mode operational space at ≤ 2.65 T and eliminate the need for He H-modes in ITER.	PFPO-2 (also affects PFPO-1 but largest impact is on PFPO-2 in which available heating is expected to provide wider operational space for H+10% He H-modes)	US1		Planned on D3D (update OS, ongoing in S21)	!
35	B.2.3	H-mode confinement with electron heating/low input torque	Evaluate H-mode confinement in ITER-like plasmas and compare with ion heated/high torque input and explore possible optimization	2	Tokamaks with appropriate heating systems to provide the required heating mix and torque input	To refine predictions of expected H-mode confinement in ITER and develop schemes for its optimization (incl. deleterious MHD avoidance/control)	FPO (this affects all ITER phases. The largest implications are on FPO as it impacts fusion power production and plasma self-heating. Specific issues for PFPO are dealt separately below),	US1		Experiments and modeling are strong on D3D. Should be extended to higher density for greater ITER relevance
36	B.2.4	H-mode access and confinement of H plasmas with electron heating and low plasma density in PFPO-1	Evaluate H-mode access and confinement in ITER-like H plasmas for PFPO with dominant electron heating	2	Tokamaks with appropriate heating systems to provide the required heating mix and low density operation in H with low core electron/ion thermal coupling	To refine predictions of expected H-mode access power and confinement in the PFPO phase	PFPO (this affects both PFPO phases because plasma density in H-mode will be restricted by the available power. The largest impact is in PFPO-1 as the only additional heating system available is ECRH)	US2		DIII-D is well positioned to study this, particularly the physics behind low density branch of L-H transition. Some of this is happening near term in the hydrogen campaign. Complementary work is ongoing in EU, in which the US is not heavily involved.	!
37	B.2.5	Pedestal parameters in H-mode plasmas with low grad-n/low n*	Determine limits to pedestal plasma parameters including transport and MHD stability for plasmas with low grad-n/low n* in H-mode plasmas	2	Tokamaks that can produce a range of density gradients in pedestal by controlling edge neutral sources	To refine H-mode pedestal plasma predictions in ITER and to determine whether pedestal transport and MHD stability will be similar - dissimilar to that in present experiments	FPO/PFPO-2/PFPO-1 (This affects all ITER phases but has largest implications on FPO as it impacts the maximum pedestal pressure and overall confinement that can be achieved. In PFPO-1 the effect may be smaller because neutral penetration in 5MA/1.8T H-modes is larger and correspondingly larger grad-n could occur in the pedestal)	US1		This topic gets at the extrapolation of pedestal to a machine with high neutral opacity and thus resilience to edge fueling. This was a major area of exploration on C-Mod, and DIII-D is now contributing strongly. It was the subject of JRT FY2019. Because of how neutral opacity scales to highest order, It will be difficult to meet the ITER condition until another high field tokamak is built and operated. Nonetheless an ITPA activity is likely to spin up on this topic, and the US should participate and can perhaps lead.
38	B.2.6	Characterization of H-mode of impurity transport in ITER-like low grad-n pedestal plasmas	Determine whether impurity density profiles become hollow in the pedestal for conditions in which temperature screening is dominant as expected in ITER pedestal plasmas	2	Tokamaks that can obtain high spatial/time resolution pedestal measurements and vary the relative gradients of n and T in the pedestal	To determine the structure of the pedestal impurity density impurity profiles in ITER and whether the assumption of dominant neoclassical transport can be used to predict impurity penetration through the pedestal in ITER or not	FPO/PFPO-2 (this affects both ITER phases but has largest implications on FPO as there will much less freedom to vary nsep by gas puffing due to the need to control divertor power fluxes in FPO when SOL powers and plasma currents will be highest.	US2		Good candidate for US leadership. C-Mod has already contributed here (Loarte, Reinke) and the actuators and diagnostics exist on DIII-D (eg LBO) to expand this. Perhaps NSTX-U could contribute something through variation of edge density gradients obtained from introducing Li. Connects with FY20 JRT.
39	B.2.7	Fuelling of He H-mode plasmas	Determine efficiency of gas fuelling of He H-mode plasmas at high edge densities and temperatures	2	Tokamaks of various sizes and PFCs that can study He H-mode plasmas over a range of edge plasma conditions	Required to determine operational density range of He H-modes in ITER (only gas fuelling is possible) and compatibility with H&CD systems	PFPO-2 (also affects PFPO-1 but largest impact is on PFPO-2 where NBI and ICRH are also available to provide much wider operational space for He H-modes)	US3		Good candidate for US leadership. C-Mod has already contributed here (Loarte, Reinke) and the actuators and diagnostics exist on DIII-D (eg LBO) to expand this. Perhaps NSTX-U could contribute something through variation of edge density gradients obtained from introducing Li. Connects with FY20 JRT.
40	B.2.8	Fuelling of H-mode plasmas by peripherally deposited pellets	Evaluate fuelling efficiency of pellets with ITER-like peripheral deposition	2	Tokamaks with HFS pellet injection that can provide peripheral pellet deposition	Required to determine efficiency of core pellet fuelling in ITER and thus of the capability to change core and edge fuelling independently	FPO (PFPO-2 is also affected but required fuelling rates are lower and thus there is more fuelling margin because H-modes will only be explored up to ~ 7.5 MA)	US2		DIII-D is already providing leadership here. Progress has been reported to ITPA.
41	B.2.9	ELM control by 3-D fields with no input torque (RF heated plasmas)	H-mode plasmas with no input torque at moderate ratios of Pinput/PLH	2	Tokamaks with in-vessel ELM control coils that can operate with RF heated plasmas and/or can control input torque	Initial H-mode operation is foreseen to be able to explore ELM control by 3-D fields in RF only heated plasmas and this may be complex due to associated mode locking if the 3-D fields are not optimized	PFPO-1 (PFPO-2 and FPO can also potentially be affected in the H-mode experiments when no NBI is applied but these are not expected to be extensive)	US2		US is a world leader in ELM control by 3D fields, via DIII-D but also collaborations. US has leadership roles in ITPA, eg Fenstermacher leading ITPA RMP group. Extrapolation to low-torque conditions is improving, both from experiment and modeling. Could do more on D3D w ECH	!
42	B.2.10	ELM control by 3-D fields with low input torque	H-mode plasmas with low input torque at moderate ratios of Pinput/PLH	2	Tokamaks with in-vessel ELM control coils that can operate with a wide range of input torques	The normalized torque from ITER H&CD systems is low and this may requires special optimization of the applied 3-D fields to avoid excessive slowing down of plasma rotation and mode locking	FPO (PFPO-2 is also affected but to a lesser degree because plasma current in H-modes is lower allowing a larger degree of 3-D field optimization)	US1		World leading efforts in this area with developments in the low torque IBS like conditions receiving a lot of attention in DIII-D Experiments. Modeling of these conditions has been increasing significantly.	!
43	B.2.11	Impurity (W) exhaust for ELM control by 3-D fields in stationary H-modes and its optimization	Determine core impurity (W) exhaust by 3-D fields and its optimization with respect to main ion particle transport	2	Divertor tokamaks equipped with in-vessel ELM control coils that can explore ELM control in a range of H-mode conditions and perform the required impurity measurements	Provides basis for impurity exhaust in ELM controlled regimes by 3-D fields in ITER at its possible optimization in PFPO for application in FPO	FPO (PFPO-1 and PFPO-2 are also impacted but to a lesser extent because the impact of W accumulation on plasma performance is less important and the risk of disruptions is lower because of the lower plasma current)	US2		LBO on DIII-D can enable such W studies combined with their existing capabilities, notably W ring campaign. Future SAS divertor with W ring can contribute. US could take a lead international role.	!
44	B.2.12	Requirements for ELM control by 3-D fields in stationary H-modes and effects on confinement and its optimization	Determine physics basis for the requirements for ELM control in ITER and quantify consequences for energy and particle confinement (thermal and fast) and possible optimization by tuning of 3-D fields to plasma conditions	2	Divertor tokamaks equipped with in-vessel ELM control coils that can explore ELM control in a range of H-mode conditions	Provides basis for the strategy to explore ELM control by 3-D fields in ITER in PFPO for application in FPO	FPO (PFPO-1 and PFPO-2 are also impacted but to a lesser extent because the impact of 3-D fields on plasma confinement has no operational consequences)	US1	C1	Have much data already; US is world leading. More analysis could be done. Also strong multi-institutional collaboration with KSTAR.	!
45	B.2.13	Control of ELM divertor power flux by mitigation	Determine relation between ELM divertor power flux/wetted area and degree of ELM mitigation/pedestal plasma parameters by ITER-like ELM control schemes (3-D fields, pellet pacing and vertical plasma oscillations)	2	Tokamaks that can mitigate ELMs with ITER-like schemes and obtain high spatial/time resolution divertor IR measurements	To determine whether ELM mitigation can provide ELM divertor power flux control or only control of the total ELM divertor energy density	FPO (PFPO-1 and PFPO-2 are also impacted but to a lesser extent because if ELM power fluxes to the wall are linked to pedestal plasma parameters these will be lower than for FPO plasmas)	US2	C2	IRTV data exist for many D3D experiments (both RMP and pellets) which could be more extensively analysed. Little or no data for vertical 'kicks'. Collaboration iwith KSTAR.	!
46	B.2.14	Impurity (W) exhaust by mitigated ELMs in ITER-like pedestal plasmas	Determine efficiency of core plasma (W) impurity exhaust by mitigated ELMs in ITER-like pedestal plasmas with low grad-n (i.e. dominant neoclassical temperature screening in the pedestal) by ITER-like ELM control schemes (3-D fields, pellet pacing and vertical plasma oscillations)	2	Tokamaks that can mitigate ELMs with ITER-like schemes and achieve ITER-like pedestal conditions (low grad-n) and perform the required impurity measurements	To determine whether ELM mitigation can provide core (W) impurity exhaust for conditions with flat or hollow impurity profiles at the pedestal.	FPO	US2		DIII-D can study this with LBO and current diagnostics. Opportunity for significant impact. Closely related to B.2.11, with similar issue re getting to high opacity.	!
47	B.2.15	Compatibility of plasma fuelling by peripheral pellet deposition and ELM suppression by 3-D fields	Optimize pellet injection (while maintaining peripheral deposition) and applied 3-D fields to achieve ELM suppression and avoid ELM triggering following pellet injection and evaluate consequences for pellet fuelling efficiency	2	Tokamaks with HFS pellet injection that can provide peripheral pellet deposition and in-vessel ELM control coils that can achieve ELM suppression for gas fuelled H-modes	To determine optimization of 3-D fields and pellet injection to provide core plasma fuelling while avoiding triggering ELMs	FPO (PFPO-2 is also affected but required fuelling rates are lower and thus there is more margin for fuelling and to optimize the 3-D fields because H-modes will only be explored up to ~ 7.5 MA)	US2		Some work done, US one of few with capabilities. Expands B.2.8. An opportunity for US leadership, particularly on DIII-D.	!
48	B.2.16	Characterization of H-mode pedestal particle transport versus particle source to establish density pedestal	Determine particle transport physics in H-mode pedestal and role of neutral source over a range of H-mode conditions	3	Tokamaks that can obtain high spatial/time resolution pedestal measurements and vary the edge neutral source	To determine effectiveness of gas and neutral recycling to fuel ITER H-mode plasmas and range of conditions over which pellet fuelling will be required	FPO/PFPO-2 (this affects both ITER phases but has largest implications on FPO as higher plasma densities and pedestal temperatures will be explored leading to lower neutral fuelling efficiency)	US2		Connected to FY19 JRT. New capabilities coming online; Ties to B.2.5. Assessing particle transport vs fueling is a focus area on DIII-D.
49	B.2.17	Effects of TF and TBM ripple on H-mode plasmas	Evaluate the effect of TF ripple and TBM ripple on H performance and its mitigation for TBMs (beyond error field correction)	3	Tokamaks that can explore the effects of TF ripple and localized TBM ripple in H-modes over a range of plasma conditions (particular edge collisionality)	Required to determine the effects over the range of TF ripple levels over which H-modes will be explored in ITER (1.3% at 1.8 T to 0.3% at 5.3 T) and the consequences of the localized TBM ripple/mitigation	PFPO-1/PFPO-2 (For TF ripple largest effects are expected at 1.8T, also can potentially affect FPO if effects of TBM ripple are found to vary strongly with Bt)	US4		Data from prior TBM mockup exist, could be analysed. No plans for new expts.
50	B.2.18	T and D transport and core DT mix control by peripheral pellet fuelling	Evaluate transport from D and T injected by pellets with ITER-like peripheral deposition and implications for DT mix control	3	Tokamaks with HFS pellet injection that can provide peripheral pellet deposition for simultaneously for two hydrogen isotopes (D and T being optimum)	Required to determine the required fuelling rates by pellet injection to control DT mix and its optimization by separate T and D pellet injection or by mixed DT pellets	FPO	US2		D3D has tools. Needs an HD pellet. Perhaps lower priority than some other topics.
51	B.2.19	First wall ELM power fluxes by mitigation	Determine level of ELM power fluxes to the first wall for mitigated ELMs by ITER-like ELM control schemes (3-D fields, pellet pacing and vertical plasma oscillations)	3	Tokamaks that can mitigate ELMs with ITER-like schemes and obtain high spatial/time resolution first wall IR measurements	To determine whether first wall fluxes for mitigated ELMs with acceptable ELM divertor power fluxes will also provide acceptable first wall power fluxes. Eventually these may be reduced by increased wall clearance (either if fluxes are excessive or if the associated Be wall erosion is excessive)	FPO (PFPO-1 and PFPO-2 are also impacted but to a lesser extent because, due to limitations in variation of nsep to maintain divertor power exhaust, pedestal W impurity profiles are more likely to be hollow)	US3		Challenging issue for D3D, or any current tokamak; would require new experiments and modifications to IR and/or use and analysis of probe array diagnostics. Would need additional effort.
52	B.3. Characterization and control of stationary power fluxes
53	B.3.1	Divertor power flux deposition width in ITER H-mode plasmas	Determine scaling of the power flux deposition width with H-mode parameters to high Bpolsep/low edge collisionality/r*, its dependence on isotope (H/D/T) /plasma specie (He) and divertor conditions	2	Tokamaks that can explore H-mode plasmas over a range of parameters (pedestal and divertor) isotopes H/D/T and species and, in particular, reach as high as possible high Bpolsep/low edge collisionality/r*, close to ITER values and can determine accurately divertor power fluxes	Determine at which point in the Research Plan control of divertor power fluxes by impurity seeding will be required to remain under the engineering limits and which gain may be expected by increasing divertor density and increasing divertor recycling	FPO (Impact on PFPO-1 and PFPO-2 is lower because power levels are lower and range of Ip in H-modes is lower)	US1	C1	D3D is contributing; strong collaboration via ITPA/DSOL. Also strong coupling with modeling efforts (eg XGC-n, BOUT, D. Curreli).
54	B.3.2	Effect of plasma response on divertor power fluxes with 3-D fields for ELM control and 3-D field optimization	Determine the effect of plasma response to 3-D fields on the spatial structure and magnitude of the toroidally asymmetric divertor power fluxes with 3-D fields for ELM control and optimization to maximize wetted area	2	Tokamaks equipped with in-vessel ELM control coils that can explore ELM control in a range of H-mode conditions and with a range of plasma responses and can perform the required divertor power flux measurements	Required to determine the spatial structure of divertor power fluxes in ITER and to identify the physics basis on which to extrapolate experimental results to ITER taking into account the plasma response expected to be required for ELM control in ITER	FPO (Impact on PFPO-1 and PFPO-2 is lower because power levels are lower and plasma densities are lower due to lower range of Ip in H-modes)	US2	C1	US-led ITPA collaborative task on this topic. Funded collaboration on KSTAR. D3D has capabilities but has been lower priority; more could be done in this topic. US also has strong modeling expertise (eg EMC3-EIRENE, MARS-F ITER Science Fellow roles of Y. Liu (GA), H. Frerichs (UW MAdison)	!
55	B.3.3	Radiative H-modes with Ne and N and mixed impurities and impact on H-mode performance	Determine physics basis to maximize divertor radiation in ITER-like plasmas and evaluate consequences for plasma performance in H-mode with ITER-like edge plasmas (pedestal collisionalities and lq)	2	Tokamaks that can explore radiative H-mode plasmas over a range of parameters (pedestal and divertor) and species and, in particular, reach as high as possible Bpolsep/low edge collisionality/r*, as close as possible to ITER values. Tokamaks should be equipped with diagnostics to accurately determine divertor power fluxes	Required to determine optimum impurity specie (or impurity mix) for efficient radiative divertor operation and to evaluate consequences of radiative divertor operation on H-mode performance	FPO (These will be the reference operating plasmas for divertor power load control in this phase due to higher power levels and Ip. These plasmas will be explored in the PFPO-2 phase but are not essential to meet the objectives)	US2	C1	Significant work at DIII-D, and collaboration with KSTAR. Could increase the connection to other programs and ITPA-DSOL. More could be done on the important topic of impurity injections; has been lower priority on D3D.
56	B.3.4	Compatibility of peripheral pellet fuelling with radiative H-modes	Determine possible limitations to radiative divertor operation due to density transients following pellet injection causing radiative collapses and optimize pellet injection and radiative H-mode conditions for integrated operation	2	Tokamaks with HFS pellet injection that can provide peripheral pellet deposition and radiative H-mode plasmas over a range of pedestal parameters	Required to determine possible limitations to H-mode radiative divertor operation caused by core fuelling by pellets in ITER or to the range of pellet sizes that can be used for core fuelling due to compatibility with H-mode radiative divertor operation	FPO (These will be the reference operating plasmas for divertor power load control in this phase due to higher power levels and Ip. These plasmas will be explored in the PFPO-2 phase but are not essential to meet the objectives)	US2		Of interest but currently no dedicated expts
57	B.3.5	Effect of 3-D fields for ELM control on divertor power fluxes in radiative H-modes	Determine the effects of 3-D fields on radiative divertor operation in H-modes for a range of plasma conditions and applied 3-D fields with a varying degree of plasma response and optimization of radiative H-mode plasmas and applied 3-D fields	2	Tokamaks equipped with in-vessel ELM control coils that can explore radiative H-modes and ELM control in a range of H-mode conditions and with a range of plasma responses and can perform the required divertor power flux measurements	Require to determine the degree to which radiative divertor operation will be effective in reducing divertor power fluxes across the divertor target in ITER and whether rigid rotation of the 3-D field structure is required to smooth off-separatrix peak fluxes	FPO (Impact on PFPO-1 and PFPO-2 is lower because power levels are lower and plasma densities are lower due to lower range of Ip in H-modes and thus radiative divertor operation with 3-D fields is not likely to be require to meet the objectives in this phase)	US2	C1	Strong US modelling activity. KSTAR and EAST collabs ongoing.	!
58	B.3.6	Wall power/particle fluxes in ITER H-mode plasmas	Determine physics mechanisms leading to wall power/particle fluxes in H-mode and their dependence on plasma edge/divertor conditions isotope (H/D/T) /plasma specie (He)/divertor geometry (vertical vs. horizontal)	3	Tokamaks that can explore H-mode plasmas over a range of parameters (pedestal and divertor) isotopes H/D/T and species and can determine accurately wall power/particle fluxes	Determines the expected level of stationary interaction of H-mode plasmas with the ITER wall and the resulting stationary power fluxes and Be wall erosion. This may eventually determine the minimum clearance between the separatrix and the wall if such fluxes are expected to be excessive	FPO (Impact on PFPO-1 and PFPO-2 is lower because power levels are lower and plasma densities are lower due to lower range of Ip in H-modes)	US3		C walls are perhaps less useful. Room for contributions, would need diagnostic add ons at DIII-D (neutral gas analysers, in-situ RGA, neutral energy analysers)
59	B.3.7	Radiative divertor operation in He H-modes	Characterize radiative divertor operation in He H-modes and compare to D plasmas including radiation control aspects	3	Tokamaks that can operate H-modes in He and D with impurity seeding and perform the required measurements of the divertor power fluxes	Required to determine whether the experience gained in developing radiative H-modes in He plasmas is relevant/useful for DT plasmas in ITER	PFPO-2	US4		He operation has been problematic on D3D due to NBI. Other tokamaks with W walls have contributed more strongly via N2 seeding expts.
60	B.4. Plasma-material/component interactions and consequences for ITER operation
61	B.4.1	Be wall erosion in He plasmas	He plasma discharges to determine if expected increased of Be wall erosion leads to coating of the divertor	2	ITER-like Be wall and W divertor is required	Required to determine if Be will coat W divertor in He plasma operation in ITER and thus avoid W-He PMI issues	PFPO-2 (also relevant for PFPO-1 but wall fluxes will be lower due to lower power fluxes and plasma current/densities)	US1	C1	In US, PISCES contributes, unique standing until FZJ has completed JULE-PSI with Be capability, ITPA task brings collaboration; needed for facility tests (esp JET).	!
62	B.4.2	He plasma modification of W mechanical properties at high fluences	Long-term exposure of W PFCs to He plasmas and characterization of exposed PFCs	2	Tokamaks with W divertor PFCs or linear plasma devices (long pulse capability is favoured to get to high fluences) operating with He plasmas (preferably H-modes to include synergies with ELMs)	Depending on results this could limit length of He plasma campaigns or eliminate them altogether. Also relevant for PFPO-2.	PFPO-2 (also relevant for PFPO-1 but divertor fluences will be lower due to lower power fluxes and plasma current/densities)	US1	C2	In US, much modelling work done in He-W system by SciDAC PSI. PISCES contributes, as will MPEX. Unknowns: He-driven effects on W recrystallization temp levels, microstructure evolution under very large He fluence conditions. Collaborations: Magnum-PSI, experiments, other facilities. ITPA task	!
63	B.4.3	Formation of fuzz by He/W interaction and critical fuzz thickness	Exposure of W PFCs to He plasmas, measurement of fuzz and impact on plasma operation	2	Tokamaks with W divertor PFCs operating with He plasmas (preferably H-modes to include synergies with ELMs)	Required to determine if fuzz is expected to grow on ITER divertor target during He H-mode plasmas and to which thickness it can grow. It should also be assessed whether such thickness is expected to affect plasma operation or not.	PFPO-2 (also relevant for PFPO-1 but divertor fuzz growth expected to be lower due to lower power fluxes/divertor temperature and He fluences)	US1	C1	In US, W fuzz layer mechanisms studied extensively, however only in linear plasma devices and single-effect labs (notably PISCES). Need studies under tokamak realistic conditions; C-Mod saw evidence of fuzz. MPEX will contribure. Will need collaborations w W devices	!
64	B.4.4	Power fluxes to castellated PFCs	Determine power fluxes to castellated structures in stationary plasmas and during ELMs over a range of conditions and identify dominant physics processes	2	Tokamaks that can expose castellated W structures to H-mode plasmas in a range of conditions and provide the necessary measurements (power fluxes, currents, etc.)	Required to evaluate power fluxes to the ITER divertor and possible melting or W material deterioration due to high surface temperatures near edges	FPO (Impact on PFPO-1 and PFPO-2 is lower because power fluxes to edges of castellations are due to lower stationary and ELM power fluxes due to lower range of Ip in H-modes)	US3	C2	PICES contributes, MPEX shuould in future, DimES at DIII-D good test facility, ITPA task. But castellated W tiles needed for full information, EU devices better equipped.
65	B.4.5	Tolerable W damage on surface and macrobrush edges for tokamak operation	Experimentally determine the tolerable level of surface damage/edge damage of divertor macrobushes to affect tokamak operation (from H-mode confinement deterioration to increased disruptivity due to uncontrolled W influxes in stationary conditions or following ELMs)	2	Tokamaks that can expose pre-damaged castellated W structures to H-mode plasmas in a range of conditions and provide the necessary measurements (power fluxes, impurity influxes, etc.)	Required to provide guidance for the tolerable W damage level for high confinement, low disruptivity (due to W influxes) H-mode operation. This may eventually limit the maximum value of the divertor power flux and/or the number of H-mode that can be performed without ELM suppression.	FPO or PFPO-2 depending on the tolerable W damage level for H-mode operation	US2	C2	US can approach using DiMES at DIIID aided by PISCES. We have some collaboration w FJZ. But, could increase activity, and modelling including connection to PMI SciDAC.	!
66	B.4.6	W operation above recrystallization and implications for tokamak operation	Determine the consequences for the W divertor material properties of sustained operation above the recrystallization temperature and assess possible synergistic effects with plasma exposure and consequences for tokamak operation	2	Tokamaks or laboratory facilities that can expose W components to plasma power/particle fluxes for sufficient lengths of time to cause significant W recrystallization while controlling component temperature (i.e. by water cooling). Tokamak experiments with water cooled components are preferred because they can also assess consequences for operation.	Required to determine an operational W recrystallization budget in ITER and thus power fluxes levels and exposure times consistent with a give degree of W surface recrystallization found compatible with appropriate tokamak operation	FPO (Impact on PFPO-1 and PFPO-2 is lower because power levels are lower, and thus power/particle fluxes, and plasma discharges are shorter)	US3	C2	Issues include effects of He and hydrogen isotope irradiation on recrystallization limits, recovery and grain growth temperature ranges over large fluence He and H plasma exposure. PISCES, MPEX, DIII-D DiMES can again contribute. But other partners have strong capabilities including W tokamaks.
67	B.4.7	W surface modification by high plasma fluence exposure and implications for tokamak operation	Determine the modification to W surface by plasma exposure to ITER-like fluences (and power fluxes, if possible) and evaluate the consequences for tokamak operation	2	Tokamaks or laboratory facilities that can expose W components to plasma power/particle fluxes for sufficient lengths of time to achieve ITER-like accumulated fluences and determine changes to W surface. Tokamak experiments are preferred because they can also assess consequences for operation	Required to determine whether there are additional limits to divertor power fluxes in ITER beyond those linked to engineering PFC limits and W recrystallization due to long term plasma exposure modification of the W surface	FPO (Impact on PFPO-1 and PFPO-2 is lower because power levels and plasma density are lower, and thus power/particle fluxes, and plasma discharges are shorter)	US3	C2	Issues include not only limits but also evolution of surface chemistry (e.g. impurity mixing and metal-impurity re-deposition) and its implications on hydrogen retention under operational conditions. PISCES, MPEX, DIII-D DiMES can again contribute. But other partners have strong capabilities including W tokamaks.
68	B.4.8	Splashing of Be and W under transients	Determine detailed physics mechanisms leading to splashing of molten Be and W PFCs under transients in tokamak experiments	2	Tokamaks that can perform controlled experiments of Be or W melting by transients and diagnose dynamics of molten material	Required to determine expected damage to Be and W PFCs in ITER under transients that cause melting and thus contribution to the determination of the required degree of transient mitigation	PFPO-2 or FPO (Impact on PFPO-1 can also be significant as ITER plasmas can generate transients that melt PFCs starting from relatively low levels of plasma current due to disruptions. For ELMs, it is expected that no melting/splashing will occur in PFPO-1. For FPO it is assumed that uncontrolled transients will be rare but plasma energies will be higher so that a few uncontrolled transients could cause significant splashing)	US4	C2	Critical issue. May be possible to do more D3D expts before openings. Gap exists in modelling, collaboration with Be and/or W expts could be increased.
69	B.4.9	Melt damage and impact on operation	Determine the impact of melt damage magnitude and spatial distribution on tokamak operation (from H-mode confinement deterioration to increased disruptivity due to uncontrolled W influxes in stationary plasmas or following ELMs)	2	Tokamak facilities that can expose pre-damaged W components to plasma discharges in a range of conditions with appropriate diagnostics of the exposed component to assess consequences for operation	Required to determine tolerable level of Be or W PFC melt damage for reliable ITER operation thus contributing to the determination of the required degree of transient mitigation	FPO and PFPO-2 (Impact on PFPO-1 is expected to be lower because the level of melt damage will be lower as well as the stationary power fluxes to molten PFCs)	US2	C2	Planned experiment on D3D w W LBO, DIMES. W PFC tokamaks required to assess consequences for operation. Collaboration via ITPA DSOL.
70	B.4.10	Dust production	Determine dominant processes for dust production from metallic PFCs by tokamak operation to provide physics basis for evaluation in ITER	2	Tokamak that can perform experiments to determine contribution to Be and W formation of plasma operation and transients with the appropriate diagnostics	Required to perform the evaluation of dust production in ITER operational phases for confirmation by experimental measurements in advance of FPO. This may impose additional transient mitigation requirements beyond those dictated by PFC lifetime considerations.	FPO or PFPO-2 (Impact on PFPO-1 can also be significant as ITER plasmas can potentially generate dust by stationary operations and transients in this phase. Depending on whether dust is produced mostly by transients or stationary operation and on the dependence of dust production on transient magnitude, dust production could be larger in the FPO campaigns or in PFPO-2)	US4	C2	Key topic but not currently v active in US. Past UCSD collaboration w AUG; may well be other collaboration opportunities.
71	B.4.11	Impact of vapour shielding on power flux in ITER transients	Determine reduction of power fluxes to PFCs under large transients due to the formation of vapour shield	2	Tokamak or laboratory facilities that can produce sufficiently energetic transients to drive the formation of vapour shields and can diagnose power fluxes to the exposed PFCs under these conditions	Required to determine expected damage to Be and W PFCs in ITER under transients for which vapour shielding is expected, thus contributing to the determination of the required degree of transient mitigation	FPO or PFPO-2 (Impact on PFPO-1 can also be significant as ITER plasmas can potentially generate transients leading to vapour shield in this phases. Due to the larger plasma energies and transient power fluxes on PFCs vapour shielding, it is expected to be larger for FPO but because the number of uncontrolled transients is lower in FPO, as mitigation schemes should operate routinely, it may be the case that the largest impact is on PFPO-2 where these schemes are developed)	US3		PISCES group has done some past measurements and modelling, but not currently active topic; could be increased. Liquid Li expts could potentially contribute. In future, NSTX-U Li experiments and US vapor box experiments will increase US contributions.
72	B.4.12	W divertor erosion under controlled ELMs	Determine the net erosion of W divertor by controlled ELMs taking into account both sputtering by the plasma (main ion and seeded impurities) and	2	Tokamaks with W divertor that can determine ELM-resolved gross erosion and net erosion after a set of well controlled experiments to make quantitative assessment,	Required to determine accumulative effects of ELMs on W divertor erosion lifetime	PFPO-1 and PFPO-2 (this impacts all phases but it is more likely to have larger impact in the PFPO phases when operation with controlled ELMs (and	was not included before, think its US3 (OS)	C2 (OS)	Update OS: Was not included before because of perception of no activities, but W-ring campaign at DIII-D has addressed this, PSI invited talk and NF paper by T. Abrams et al., and more), therefore I cincluded this
73	B.5. Start-up, Ohmic, L-mode scenario development
74	B.5.1	ECRH assisted plasma start-up	Perform experiments to optimize ECRH assisted start-up and to validate models for ITER	2	Tokamaks that can perform ECRH assisted start-up over a range of conditions (e.g. applied electric field, etc.) and can diagnose the plasma in this initial phase	Required to design experimental strategy to optimize plasma start-up in ITER and minimize experimental time dedicated to it	FP and PFPO-1 because these are the first phases in which ECRH assisted start-up will be used	US1		D3D is planning EC startup experiments. (one of few current). Others have been done in past, we are providing input to models.	!
75	B.5.2	Ohmic plasma start-up	Perform experiments to achieve ohmic plasma start up in ITER-like conditions (i.e. electric field) and identify the range of parameters over which this can be most reliably achieved Optimization of experimental strategy to achieve Ohmic start-up by tuning of PF operation. Validation of models to describe plasma transport in initial ohmic phase and tokamak electromagnetic model used for ITER PF optimization	2	Tokamaks that can perform ohmic assisted start-up over a range of conditions (e.g. applied electric field, etc.) and PFCs, including Be wall, and can diagnose the plasma in this initial phase.	Required for an initial evaluation of the need for ECRH assisted plasma start up for 1.8 T plasma operation	PFPO-1 (also for PFPO-2 as 1.8T operation is foreseen to compare with reference PFPO-1 plasmas without TBM)	US3		Much existing data from all devices including C-Mod, NSTX and D3D.
76	B.5.3	Need for central heating to control W in L-mode	Determine the required level of central heating to avoid W accumulation in L-mode plasmas versus divertor plasma conditions	3	Tokamaks with W PFCs that can vary the level of central heating in L-mode power while spanning a range of divertor conditions	Required to determine optimum path from low Ip to high Ip L-modes, as central heating levels in ITER are Bt dependent for ECRH and ICRH and central heating is not maintained for all paths	PFPO-2 (FPO will be impacted in a similar way as the optimum path in L-mode is expected to be developed in PFPO-2 for H plasmas and tuned for DD and DT plasmas in FPO)	US3		expts planned on D3D
77	B.5.4	Plasma transport in ramp-up phase	Determine density, temperature and plasma transport in the ramp-up phase of ohmic and L-mode heated plasmas to validate models for ITER	3	Tokamaks that can explore various levels of additional heating and ramp-rates and determine temperature and density evolution and plasma transport in the ramp-up phase	Required to optimize current profile evolution during the ramp-up phase of ITER scenarios by tuning of density, heating power and current ramp-rate to achieve suitable q profile at the end of the ramp (for long pulse Q = 5 scenarios) or to reduce flux consumption in ramp-up (for Q = 10 scenario)	FPO (PFPO-1 and PFPO-2 can also be affected but the degree of optimization required for ITER plasma scenarios is much higher in FPO)	US3		DIII-D and NSTX-U focus on ramp-up optimization. Need to extend analysis and modeling to ITER regimes.
78		ECRH Stray radiation: avoidacne and characterization of absorption						?		Details needed to comment
79	B.6. Conditioning, fuel inventory control
80	B.6.1	ECWC conditioning	Perform ECWC conditioning and determine requirements for effective cleaning and reliable plasma start-up including recovery from unmitigated and mitigated disruptions	1	Tokamaks that are equipped with ECRH systems and can perform ECWC on a routine basis for general machine conditioning and disruption recovery and diagnose the ECWC plasma	This is one of the two cleaning technique available in ITER with toroidal field on. This is the only cleaning technique for PFPO-1 with the TF field on. Without an effective cleaning technique in this phase the experimental programme could be significantly slowed down (e.g. recovery from disruptions or disruption mitigation)	PFPO-1 because ECWC is the only conditioning technique possible with the TF on in this phase	US3		ITER has requested D3D do this. But can only be short pulse - SS tokamaks seem to us better suited.
81	B.6.2	Operation at low wall temperatures and ITER-like baking cycles	Determine resulting fuel retention by operation at ITER-like temperatures during non-active phases and compare to higher wall temperature operation	2	Tokamaks with ITER-like wall/divertor materials and capable to operate at ITER-like temperatures for wall and baking and to measure outgassed and retain fuel	Required to determine if the wall temperature plays a major role in ITER fuel retention as wall temperature will be different between PFPO and FPO and this could affect the evaluation of expected T retention in FPO on the basis of PFPO measurements in ITER	PFPO-1 and PFPO-2	US5	US2	needs relevant PFCs. US is leading ITER RGA development, collaborating on other tokamaks.
82	B.6.3	Removal of hydrogen by baking and GDC	Determine effectiveness of ITER pre-operational campaign conditioning cycles based on baking and GDC for a range of glow conditions	3	Tokamaks that can perform similar conditioning cycle as foreseen in ITER including plasma facing materials and baking temperature	Required to optimize the pre-operation plasma conditioning cycle and the of the GDC glow	All phases, as this is the conditioning sequence foreseen before starting of plasma operation in all phases	US5		This and rest of section seem to require tokamaks with W+Be. Some possibility of international collaboration, but conditioning is typically best studied by local operating team.
83	B.6.4	Efficiency of T removal by plasma operation from Be codeposits in routinely unexposed areas	Evaluate the efficiency of fuel removal in divertor Be codeposits (by isotopic exchange and/or thermal desorption by plasma heating) not routinely exposed to large plasma particle/power fluxes by location of the plasma separatrix in these locations.	3	Tokamaks with W/Be PFCs where Be codeposits in remote area can be accessed by the plasma with suitable modifications to the magnetic configuration	Required to refine the T housekeeping strategy in ITER. If effective this could be adopted for routine ITER cleaning pulses (operation with raised strike points on the Be co-deposits) avoiding the need of more complex T removal techniques.	FPO	US5		see above
84	B.6.5	ICWC fuel removal from Be codeposits	Evaluate the efficiency of fuel removal in divertor Be codeposits by ICWC	3	Tokamaks that can apply ICWC on Be codeposits with a range of thicknesses and can diagnose the resulting outgassing	Required to refine the T housekeeping strategy in ITER, It is required to determine the need and frequency of ICWC to maintain a low level of in-vessel T retention as removal effectiveness can depend on characteristics of Be codeposits.	FPO	US5		see above
85	B.6.6	Start of operation by baking without GDC	Determine if it is possible to start plasma operation following a machine opening to air by only baking in tokamaks with metallic PFCs	3	Tokamaks with metallic PFCs (W/Be preferred) that can perform ITER-like baking cycles	Required to determine to which degree GDC is mandatory to restart plasma operation in ITER given possible issues of GDC electrode lifetime	FPO	US5		see above
86	B.7. Basic scenario control and commissioning of control systems
87	B.7.1	Development of criteria for allowable n = 1 and n =2 error fields in ITER (locked mode threshold, increased disruptivity in H-L transitions, etc.)	Development of scaling of critical n =1 and n = 2 error fields for ITER operation	2	Tokamaks that can apply error fields in a controlled way (i.e. by external error field coils) from various locations in the plasma cross section (LFS, HFS, Top/Bottom)	To provide criteria on acceptable deviations of TF, CS. PF coils from their ideal shape and position due to manufacturing and positioning tolerances. To evaluate the requirements to correct error fields caused by ferromagnetic elements in ITER. To optimize error field correction in ITER by the use of external coils, possibly supported by internal coils.	PFPO-1, PFPO-2 (addition of TBMs) and FPO (larger plasma bN)	US1		D3D a leader in such experiments.	!
88	B.7.2	Noise in dZ/dt and development of means to reduce it	Determine the source of noise in dZ/dt (plasma/hardware) signal used for VS control and develop schemes for its minimization that can be ported to ITER	2	A range of tokamaks to compare source of noise in dZ/dt and determine its origin as well as to demonstrate that noise reduction techniques are robust across devices	Required to minimize AC losses in the superconducting coils driven by reaction of PF coil system on noise in dZ/dt to keep VS stability. Required to reduce heating of the VS in-vessel coils/busbars and, possibly, to reduce thermal fatigue of the divertor due to strike point oscillations.	FPO	US2	C1	U.S. machines have done work on this, but can continue to 1) explore sources of noise in vertical position/velocity estimates and 2) explore how filtering methods improve control (will reduce noise sensitivity but possibly add lag and reduce overall stability).
89	B.7.3	Error field identification and correction	Determination of schemes to optimize error field identification and correction on the basis of the as-built tokamak component specifications and for a range of plasma scenarios	3	Tokamak with ITER-like systems to correct (external coils) and to identify error fields (in-vessel coils) that can explore error field identification and correction for a similar range of plasma conditions to those in ITER	To optimize the strategy for error field identification and correction in the various phases of the research pan	PFPO-1, PFPO-2 (addition of TBMs) and FPO (larger plasma bN)	US1		NSTX-U had an active research program on this during 2016 campaign and will as soon as the machine starts up (will be a priority since there have been so many changes to the device).	!
90	B.7.4	Impact of application of 3-D fields for ELM control on plasma position control	Determine consequence of the application of 3-D fields for ELM control on plasma position control with ITER-like sensors and develop schemes for optimum position control	3	Tokamaks equipped with in-vessel coils for ELM control and magnetic sensors with similar distribution to that of ITER	Required to optimize plasma position control with applied 3-D fields for ELM control	FPO (this also affects PFPO-1 and PFPO-2 when 3-D fields are applied for ELM control, but for FPO the consequences of a plasma position error are much larger because of the larger plasma energies and power fluxes)	US2		D3D has strong capabilities. Would benefit from modelling that could be tested on experiments.
91	B.7.5	Develop real time ICRH coupling control loop	Develop a real time ICRH coupling control loop with ITER-like actuators (e.g. separatrix position control) and (normalized) timescales for actuators and plasma portable across tokamaks	3	Tokamaks with ICRH heating and suitable ITER-like actuators that can implement such control loop	Required to ensure required ICRH power through time-varying plasma conditions	PFPO-2 and FPO	US4		EU leading. Little current US experimental work on ICRF, but some on other actuators on D3D and NSTX. Some collaboration w JET.
92	B.7.6	Develop real time divertor power flux measurement and control loop	Develop a real time divertor power flux measurement and control loop with ITER actuators (power, gas fuelling, impurity seeding), sensors, and (normalized) timescales for actuators and plasma portable across tokamaks	3	Tokamaks with good divertor power flux diagnostics and other ITER-like sensors that can be used to provide a real time measurement and can implement such control loop	Required to ensure a given level or upper value) of divertor power flux through time-varying plasma conditions	FPO (PFPO-1 and PFPO-2 will also be affected but power fluxes are expected to be much lower so that control loop can be tested but is not required in these phases)	US2		D3D doing relevant work. C-Mod has published relevant work. NSTX-U plans real time modelling.
93	B.7.7	Develop density control loop based on gas and pellet fuelling	Develop a real time plasma density control scheme based on gas and pellet fuelling with ITER-like (normalized) timescales for actuators and plasma portable across tokamaks	3	Tokamaks with gas and pellet fuelling systems that can implement such control loop	Required to ensure good density control in stationary and transient phases	PFPO-1 and PFPO-2 as this control loop needs to be developed before FPO	US2		D3D has tools, could increase emphasis on routine, ITER-like control. Currently more active on gas vs pellet fueling. Would benefit from increased control simulations to benchmark - US modelers could also contribute.
94	B.7.8	Optimize NTM control commissioning	Define an experimental strategy to commission NTM control that can be ported across tokamaks and minimizes the number of plasma pulses/conditions required	3	Tokamaks capable to stabilize NTMs with ECRH/ECCD over a range of experimental conditions	Required to minimize experimental time dedicated to NTM control commissioning	PFPO-2 and FPO	US2		Ongoing work on control; more work is needed on strategy
95	B.7.9	Optimize Sawtooth control commissioning	Define an experimental strategy to commission sawtooth control that can be ported across tokamaks by minimizing the number of plasma pulses/conditions required	3	Tokamaks capable to stabilize sawteeth by ECRH/ECCD and ICRH/ICCD over a range of experimental conditions	Required to minimize experimental time dedicated to sawtooth control commissioning	PFPO-2 and FPO	US2		Some current efforts but more focussed effort could be done.
96	B.7.10	Installation and demonstration of PCS in a tokamak with ITER-like actuators/timescales	Install and demonstrate plasma operation with PCS in a tokamak by suitable tuning of the actuators to mimic ITER-like operation	3	Tokamaks with ITER-like actuators that can replace their control system by PCS	Required to optimize testing and refinement of PCS for ITER operation by application to real tokamak operation	PFPO-1 as PCS needs to be ready for commission and operation for initial plasma operation	US2		AUG and KSTAR are doing components of this. D3D could provide unique and critical demonstrations of some of the key algorithms of the ITER PCS in much more high fidelity ITER plasmas, as well as using its PCS to scale the dynamics of the plasma and having the RTF control with the same dynamics (i.e. exactly the SAME algorithms) as ITER itself. Large task which would require additional resources.
97	B.8. Transient phases of scenarios and control
98	B.8.1	ELM control and W accumulation control in L-H and H-L phases at constant and varying Ip	Demonstrate ITER-like ELM control schemes (3-D fields, pellet pacing) during H-mode access and exit phases with constant and varying Ip	2	Tokamaks with ITER-like ELM control schemes (3-D fields from in-vessel coils and pellet pacing) that can apply them to a range of H-mode scenario access/exit phases. For assessment of W accumulation control, central heating capabilities and W PFCs are required	Required to ensure robust entry to and exit from stationary H-modes while maintaining ELM control and avoiding W accumulation both when entry/exit takes place during the current flat top as well as with evolving Ip/q95	FPO (PFPO-1 and PFPO-2 are also affected but consequences of lack of ELM control/W accumulation in these phases are largest for FPO due to the larger plasma energies/power/current levels)	US2		D3D is working on ELM control, and doing impurity control expts using laser blowoff. Could be increased to integrate these various aspects, more on L-H and H-L. AUG likely leading.	!
99	B.8.2	Development of integrated H-mode termination scenarios	Demonstrate integrated H-mode scenarios with controlled density evolution, plasma radiation, divertor power fluxes, etc. by actuators available in ITER and with relevant normalized timescales	2	Tokamaks with ITER-like actuators (H&CD, fuelling, impurity seeding, etc.) that can apply them to a range of H-mode scenario termination phases and demonstrate control of required parameters (suitable measurements are required)	Required to ensure robust exit from stationary H-modes while avoiding plasma physics limits (e.g. density limits) and operational limits (e.g. excessive divertor power fluxes)	FPO (PFPO-1 and PFPO-2 are also affected but consequences of lack of ELM control/W accumulation in H-mode termination are largest for FPO due to the larger plasma energies/power/current levels)	US1		D3D has some unique features, including low torque. Connection to modelling.
100	B.8.3	Dynamic error field correction for transient confinement phases	Develop dynamic error field correction schemes for transient confinement phases (L-H, H-L) portable across tokamaks to mitigate error field effects on confinement and minimize risks of disruptions in these phases	3	Tokamaks equipped with in-vessel coils that can apply time varying error field correction within the timescale of confinement transient phases	Error field correction depends on plasma conditions. Transient confinement phases can be sensitive to error fields particularly in ITER in which plasma rotation is expected to be low	FPO (this also affects PFPO-1 and PFPO-2 but error field effects are expected to be larger with larger bN so that a larger level of correction is required in FPO)	US2	C2	More expts and algorithm development on US machines could contribute strongly. Collaboration with international devices would allow comparions of different types of error field, coils, plasma parameters etc.