ITER-Research-Plan_final_ITR_FINAL-Cover_High-Res

O BJECTIVES

The ITER project aims to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes, as outlined in the Project Specification [ITER_D_2DY7NG, 2010] To achieve this overarching objective, a series of scientific and technical goals related to plasma fusion performance have been established.

The device must facilitate prolonged burn in inductively driven plasmas, achieving a fusion power to auxiliary heating power ratio (Q) of at least 10 (Q ≥ 10) This capability should be demonstrated across various operating scenarios and maintained for a duration adequate to establish stationary conditions consistent with the typical timescales 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 a fusion 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 ITER Research Plan (IRP) aims to outline the necessary research, development, and facility exploitation to achieve ITER's mission goals This program encompasses activities related to defining performance requirements for ITER components and includes R&D from other facilities to establish a robust physics and technology foundation for ITER's efficient operation The IRP will ultimately detail the comprehensive range of R&D activities necessary to meet specific scientific and technical objectives While previous versions focused on the research program to fulfill key scientific goals, discussions on technology activities have primarily centered around the Test Blanket Module (TBM) testing program Future iterations of the IRP should expand the scope of technology R&D to cover extensive research on the ITER tokamak and its auxiliary and plant systems, addressing the full spectrum of the project's technical mission goals.

The IRP analyzes the ITER facility's capacity to fulfill its mission and explores necessary upgrades to tackle science and technology challenges related to specific mission objectives The current version evaluates various upgrade options and their potential effects on the scientific research program during the Operations Phase Additionally, a risk analysis has been conducted to assess the likelihood and consequences of unfavorable R&D or exploitation outcomes, identifying possible risk mitigation and avoidance strategies where applicable.

The Research Plan is structured around key thematic research areas that collectively create an exploitation plan for the facility This plan is incorporated into the Staged Approach strategy established during the 2015 revision of the ITER project schedule.

In 2016, the strategic scheduling approach significantly influenced the overall structure of the Integrated Research Program (IRP) and the timing of essential activities, particularly the transition to operational use of deuterium-tritium fuel, which is crucial for achieving substantial fusion power production.

The experimental program within the IRP has outlined estimated durations for its various elements; however, these estimates carry uncertainty, especially as the project advances towards fusion power production This involves the study and optimization of burning plasmas, a novel area of laboratory plasma research that ITER will facilitate At this stage, a critical outcome of the plan is the logical sequence of programmatic activities and the identification of essential prerequisites for implementing major activities, including the availability of auxiliary hardware and the outputs from prior steps in the plan.

As the project transitions into the Operations Phase, it is crucial to regularly review the ITER Research Plan (IRP) to incorporate findings from ongoing research and development, as well as insights gained from ITER's experimental operations The IRP should be regarded as a dynamic document that adapts to the project's progress Furthermore, it acts as a baseline to assess the impact of design modifications, planning adjustments, or unexpected occurrences on the successful attainment of the ITER mission.

M ETHODOLOGY

The Initial Research Plan (IRP-v1.0), presented to the ITER Council Science and Technology Advisory Committee (STAC) in May 2008, was developed during the 2007-2008 Design Review to guide the scientific research program essential for achieving the project's key scientific mission goals However, changes in the project schedule, particularly regarding the hardware configuration for First Plasma and the installation of an all-tungsten divertor, required significant revisions to the IRP Several updates were made to align with these changes, culminating in IRP-v2.1, which was presented to STAC in October 2009 and later incorporated into the ITER Baseline as IRP-v2.2 in July 2010 Despite a further update, IRP-v2.3, being prepared for STAC in May 2010 to reflect the approved ITER Project Schedule targeting First Plasma in late 2019, it did not complete the internal review process in time for the 2010 Baseline, leaving IRP-v2.2 as the current reference version of the Research Plan.

The Research Plan has undergone significant revisions since 2010 to accommodate updates in the project schedule and hardware configuration, notably the choice to commence the experimental program using an all-tungsten divertor, amid ongoing uncertainties regarding delivery timelines.

Since 2010, 'deferred' components have hindered the development of a fully self-consistent revision of the Research Plan However, the comprehensive revision of the ITER project schedule conducted between 2015 and 2016 laid the groundwork for an updated Integrated Research Plan (IRP), aligning it with the anticipated ITER tokamak and facility configuration from First Plasma to full deuterium-tritium (DT) operation.

The revised Integrated Research Plan (IRP) was created in collaboration with representatives from the Members' fusion communities, nominated by Domestic Agencies (DAs), who provided extensive expertise in fusion physics and tokamak operations To analyze the proposed experimental program for the ITER Operations Phase, three dedicated workshops were held at ITER Headquarters from July 2016 to March 2017, involving the ITER Organization Central Team (IO-CT) and fusion community experts.

The workshops facilitated in-depth discussions between IO-CT staff and fusion community experts regarding the experimental program's structure within the Research Plan They addressed the specific experimental activities to be undertaken, associated risks, and necessary preparatory R&D Additionally, the sessions provided a chance to evaluate the estimated experimental time needed for the proposed program based on current operational experiences from existing fusion facilities.

During the workshop, participants received a summary of the outputs from the Configuration Workshop held from June 27 to July 1, 2016, where the facility hardware configuration was developed by IO-CT and IO-DA management and staff Additionally, the workshop provided essential background information on key activities and issues related to the existing Research Plan, referencing the most recent report presented to STAC in 2014.

Prior to the workshop, participants received a report that served as a key reference for their analysis, along with supplementary information on experimental risks from previous Research Plan analyses Detailed insights into the hardware configuration of essential ancillary systems and critical ITER physics were shared through introductory presentations To enhance the examination of significant activities within the Research Plan, participants were organized into various Working Groups, each tasked with focusing on specific components of the experimental program.

The Working Groups were tasked with analyzing essential activities for developing the research program post-First Plasma, ensuring alignment between experimental activities and the planned hardware configuration within the Staged Approach This includes commissioning all necessary auxiliary systems to full performance with suitable pulse durations, developing the experimental capabilities for initial DT fusion power production at several hundred MW, and proposing a comprehensive experimental program in DT plasmas to achieve key mission goals in both inductive and non-inductive scenarios, while also addressing the requirements of the TBM testing program.

 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;

 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 the elements 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 outlined in this document comprehensively details the progression of the scientific research and TBM testing programs throughout the four phases of the Staged Approach Additionally, it expands the analysis to include developments beyond the initial phase of DT operation, focusing on the advancement of long-pulse, high fusion power plasmas that align with the project's primary mission objectives.

O VERALL S TRUCTURE OF THE ITER R ESEARCH P LAN P RESENT V ERSION

ITER R ESEARCH AND O PERATIONAL P HASES

The Research Plan has been revised to align with the Staged Approach strategy in the updated ITER schedule, aiming for a transition to nuclear operation by late 2035, following First Plasma in late 2025 Additionally, the experimental program for DT plasmas will be expanded to include the development of long-pulse plasmas that achieve high fusion power and gain, in accordance with the project's mission objectives.

The revised operational schedule consists of four stages, culminating in an extensive development program focused on DT fusion power production during long-pulse operations.

DT scenarios and a research program in burning plasma studies), the principal goals are:

The achievement of the First Plasma milestone marks a significant step in plasma breakdown using hydrogen or helium, reaching a plasma current of at least 100 kA for a duration of 100 ms Following this milestone, the Engineering Operation phase will focus on commissioning the magnet systems to full current, with potential attempts to achieve circular limiter plasmas of up to 1 MA.

Pre-Fusion Power Operation 1 (PFPO-1) will involve comprehensive commissioning with plasma activities to achieve diverted plasma operation in hydrogen/helium, targeting at least 7.5 MA/2.65 T This will be supported by plasma control, diagnostics, and commissioning of Electron Cyclotron Resonance Heating (ECRH) and Ion Cyclotron Resonance Heating (ICRF), assuming the availability of a 10 MW ICRF heating antenna Additionally, a systematic disruption management and mitigation capability will be developed for higher current operations An option for operating at a toroidal field of 1.8 T is under analysis, potentially enabling early access to H-mode operation with up to 30 MW of ECRH and ICRF power This strategy may involve implementing dual-frequency operation of gyrotrons or dedicating some to lower frequency operation, necessitating modifications to the Electron Cyclotron transmission and launching systems for plasma start-up and heating.

The Pre-Fusion Power Operation 2 (PFPO-2) will prioritize the commissioning of the Heating Neutral Beam (HNB), Diagnostic Neutral Beam (DNB), and ICRF systems to full power, along with essential plasma control capabilities The program aims to develop H-mode operation in hydrogen, with helium as a backup, focusing on reliable ELM control and divertor heat flux management A significant objective is to achieve high power L-mode operation to showcase the device's full technical potential Systematic investigations into plasma-wall interactions, including issues like erosion, redeposition, dust production, fuel retention, and removal, will also be conducted in preparation for the nuclear phase If 1.8 T operation is implemented, H and He H-modes could be explored, providing critical data for comparing H-modes at varying fields Furthermore, if H-mode operation at 1.8 T was successfully achieved in PFPO-1, it would facilitate the assessment of the 3-D fields generated by the TBMs, allowing for the development of necessary mitigation strategies.

The Fusion Power Operation (FPO) campaign aims to leverage previous experimental capabilities to demonstrate high power H-modes in deuterium, setting the stage for deuterium-tritium (DT) operations Initially, trace tritium experiments will lead to a gradual transition to full DT operation as the Tritium Plant's fuel reprocessing capacity increases The primary objective is to achieve fusion power production of several hundred megawatts for several tens of seconds, targeting a Q value between 5 and 10 Future experimental campaigns will focus on optimizing fusion gain to reach Q values of 10 or higher, extending pulse durations in inductive plasma scenarios towards 300 to 500 seconds, and developing hybrid and fully non-inductive scenarios to achieve burn durations of 1000 to 3000 seconds with a Q value around 5.

The exploitation phases of the research plan are centered around key issues that guide an experimental program aimed at commissioning plasma systems, developing control capabilities, and establishing specific plasma scenarios This initiative addresses uncertainties in plasma physics at the ITER scale and in the burning plasma regime, while also conducting in-depth studies to optimize fusion performance However, there is increased uncertainty regarding the later phases of the Research Plan, as earlier phase results will significantly influence the planning of subsequent phases, making ITER the pioneering experiment for studying burning plasmas.

This document outlines the scientific research program during the Operations Phase, specifically detailing the Test Blanket Module (TBM) Testing Program It encompasses the initial operational tests of TBMs in PFPO-2 and extends to the completion of the first integrated tests, focusing on the operational behavior of the blanket component for heat extraction and tritium breeding/management, anticipated to occur 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 in parallel to the construction phase These activities are intended to complete the design basis for certain key auxiliary systems, to resolve issues impacting the efficiency of ITER operation, to prepare plasma scenarios to be used in ITER and to develop specific experimental techniques and tools 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 the development of the Research Plan

The current version of the IRP includes several appendices that address key issues related to the research program's implementation Appendix A focuses on deuterium 'spiking' in hydrogen plasmas during PFPO-2 to enhance fuel retention estimates Appendix C examines heat load management for the ITER divertor, while Appendix D outlines a routine 'reference pulse' for monitoring plasma-facing materials' conditions Appendix B reviews the experimental implications of H-mode power threshold scaling for ITER and the effects of TF ripple on H-mode performance Additionally, Appendix F discusses the physics considerations for H-mode operation at 1.8 T, a recently proposed option for the research program.

The revised ITER schedule proposes the elimination of a later inclusion of an LHCD system, prompting a review of modeling studies on long-pulse and non-inductive DT scenarios This review compares predicted plasma performance with and without LHCD in the H&CD upgrade options, focusing on the Q-value and proximity to a fully non-inductive state Additionally, plans for the phased installation of baseline H&CD systems, potential modifications to the installation schedule, and the implications of various upgrades are discussed The strategy for the phased installation of baseline diagnostic capabilities is outlined, while considerations for tritium availability from external sources to support the DT experimental program from the late 2030s to early 2050s are summarized Lastly, an updated Research Plan Risk Register highlights significant risks to the successful implementation of the experimental program and the timely achievement of fusion power production.

A SSUMPTIONS ON THE P HASING OF ITER O PERATIONAL C APABILITIES

The revised ITER schedule under the Staged Approach focuses on the phased installation of the operational capabilities of the ITER tokamak and facility A comprehensive discussion of the proposed Plant Configuration for each phase was presented to the STAC in 2016, and the baseline configuration has been formalized in 2017 Figure 1.3-1 illustrates the evolution of the plant configuration during the Staged Approach, highlighting options for PFPO-1 that are under evaluation Key elements influencing the Research Plan implementation are summarized for reference, with further details on the phasing of H&CD systems and diagnostic installation sequences available in Appendices G and H The evolution of auxiliary systems has been agreed upon to support the experimental program outlined in sections 2.4, 2.5, and 2.6.

 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 the plasma)

 A limited subset of the baseline Diagnostic capability, essentially to characterize plasma breakdown and to provide limited plasma control and machine protection

 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 and machine 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

The experimental program will necessitate the support of all listed systems, and additional options are currently under investigation If these options prove feasible and beneficial to the research program, they may be implemented through future Project Change Requests (PCR).

To enhance plasma start-up and heating, it is essential to incorporate up to one-third of the installed ECRH power with low frequency capabilities, utilizing either eight dual-frequency gyrotrons operating at 104 GHz or eight dedicated low frequency gyrotrons at 110 GHz.

 Acceleration of one ICRF antenna to provide up to 10 MW injected power capability at

 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 be available

The ITER 'Plant Configuration' schematic illustrates the various phases of the Staged Approach, highlighting when each system becomes fully operational, as indicated by the color code Temporary connections for the delivery and exhaust of non-tritiated gases during the PFPO phases are also noted.

(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

 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):

During the PFPO phase, it is crucial to ensure that all elements of the tokamak and its auxiliary systems are operational to facilitate the efficient execution of the D and DT phases of the program Key advancements for the nuclear phase will include the capability to measure 14 MeV neutrons and various fusion product diagnostics, scheduled before the nuclear operations commence A primary objective of the PFPO phase is to fully commission the Heating and Current Drive (H&CD) systems at least 50 seconds prior to the nuclear phase, while ensuring all fueling systems are operational with all necessary gases, excluding deuterium and tritium Additionally, the PFPO phase aims to establish effective magnetic and kinetic control, demonstrate reliable disruption mitigation, and achieve ELM suppression By the onset of the nuclear phase, all plant and ancillary systems should be in an advanced operational state Concurrently, the Tritium Plant's processing throughput will expand gradually during deuterium experiments and into the initial DT operation period If necessary upgrades to meet the Q = 10 goal are identified during PFPO, they are likely to be implemented early in the FPO phase, requiring dedicated commissioning time.

K EY R ESEARCH I SSUES

1.3.3.1 Key Research Issues during Operation

The Research Plan for the exploitation phases focuses on a series of 'commissioning with plasma' activities and essential research issues While some issues will be tackled simultaneously to optimize experimental time, many must be resolved before advancing to subsequent challenges.

The primary physics objective of the PFPO is to develop plasma scenarios for system commissioning, focusing on heating and current drive systems This process requires extensive commissioning of the plasma control and protection systems, including the Central Interlock System (CIS), Central Safety System (CSS), and Plasma Control System (PCS), to ensure the safe and reliable operation of the tokamak Key activities involve utilizing in-vessel coils for vertical stabilization of highly shaped plasmas, studying error fields in ITER, and assessing their impact on plasmas, along with routine corrections using dedicated coil sets Concurrently, essential research will focus on characterizing disruptions, developing reliable disruption prediction methods, and optimizing the disruption mitigation system (DMS) as plasma parameters gradually increase Additionally, early commissioning of diagnostic systems and data validation is crucial for enhancing measurement capabilities for plasma control, disruption detection, and machine protection, particularly for all-metal plasma-facing components before significant heating power and higher plasma thermal energies are introduced.

Once adequate measurement, control, and protection capabilities are established, the commissioning of Heating and Current Drive (H&CD) systems will be prioritized to enhance plasma parameters and confinement Initial physics studies during this phase will provide the first opportunity to investigate tokamak plasma behavior at the ITER scale Although uncertainties remain regarding the H-mode power threshold scaling to ITER, experimental data suggests that H-modes may be achievable in helium plasmas using ECRH heating alone Operating ITER at 1.8 T with an enhanced H&CD capability, including an additional 10 MW of ICRF, could facilitate early studies of hydrogenic H-modes, previously deemed unlikely in this initial operational phase Access to H-modes would enable early evaluations of critical issues such as ELM control and aid in refining the requirements for future heating and current drive upgrades Additionally, important plasma-wall interaction topics will be of interest, even within the limited plasma parameter range available during this period.

The upcoming installation of enhanced Heating & Current Drive (H&CD) and Diagnostics capabilities prior to PFPO-2 will significantly broaden the scope of plasma scenarios and parameters for research, enhancing ITER's physics studies During this phase, 'commissioning with plasma' activities will be crucial, focusing on the testing of tokamak and plant systems ahead of the D/DT operational transition Key areas of exploration will include advanced plasma control techniques such as NTM suppression, sawtooth control, and divertor heat flux management, alongside a comprehensive program for H-mode studies, particularly ELM control The experimental campaign aims to extend the plasma operating range to high currents and long pulses, with initial operations expected at 15 MA/5.3 T in L-mode Additionally, the current drive capabilities of HNB systems may enable extended pulse lengths in both L- and H-modes, contingent on successful commissioning Initial investigations into current drive efficiency and target q-profile formation for hybrid and non-inductive scenarios will also be initiated Furthermore, Power-Wall Interactions (PWI) studies will play a vital role, focusing on power deposition, erosion, material migration, fuel retention, and dust generation, to develop a comprehensive understanding of these processes prior to DT operation.

Following the transition to the FPO phase, plasma operations will focus on deuterium experiments, which are essential for developing H-mode scenarios for future DT use The duration of the deuterium program will be influenced by the Tritium Plant's commissioning pace, as it will start with limited throughput that will increase over time While deuterium L- and H-mode studies will allow for extensive exploration of plasma behavior at the ITER scale, the program aims to transition to 50:50 DT plasmas quickly once adequate tritium supply is available Initial experiments using trace tritium in deuterium plasmas will be valuable for understanding tritium fueling and transport processes If the Tritium Plant's throughput increases slowly, more time may be dedicated to deuterium scenario development to reduce overall development time.

DT During this program, control of ELMs, core MHD and (simulated) plasma burn will be established and combined into fully integrated scenarios

The initial phase of the ITER Research Plan will concentrate on experimental R&D in conventional inductive scenarios while also beginning the development of advanced and hybrid scenarios, contingent on the available operational time and tritium throughput The primary goal is to achieve full deuterium-tritium (DT) operation as quickly as possible, prioritizing the establishment of a solid foundation for significant fusion power production and gain in the most promising plasma scenario, specifically the inductive ELMy H-mode, for durations of several tens of seconds The extensive development of more advanced scenarios will be postponed until later in the DT phase.

Once a sufficient tritium throughput is achieved, the transition from deuterium to full DT plasmas will involve gradually increasing both plasma current and tritium fraction This approach allows for the integration of critical factors such as power handling, burn control, fuel mixture control, and helium exhaust, initially established during the deuterium phase The primary objective is to develop an integrated scenario that can generate several hundred megawatts of fusion power for several tens of seconds Successfully accomplishing this during the first FPO campaign will serve as a significant demonstration of ITER's capabilities and the potential of magnetic fusion energy.

Once fusion performance reaches the desired level, the program will focus on optimizing DT scenarios to demonstrate a Q-value of 10, initially achievable in short pulses lasting several tens of seconds Concurrently, a comprehensive study of burning plasma behavior in the most effective scenarios will pave the way for advanced fusion plasma research Following this, the enhancement of plasma performance and the investigation of burning plasma physics will be integrated into a strategy aimed at optimizing Q and extending pulse durations to several hundred seconds Additionally, experimental efforts will prioritize the development of hybrid (or improved H-mode) and fully non-inductive scenarios Achieving long-duration burns with a Q-value nearing 5 in the latter scenario will likely require several years of experimentation, while the hybrid mode could provide a faster path to long-pulse operation if current experimental findings translate effectively to ITER's scale and burning plasma conditions.

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;

 Physics of the plasma-boundary interface;

 Integrated burning plasma scenario development

Advanced modes in fusion research face critical challenges such as current profile control and magnetohydrodynamic (MHD) control at pressures exceeding the no-wall β-limit Addressing these issues is essential for fulfilling the ITER core mission and maximizing the unique advantages ITER offers to the global fusion program Notable capabilities of ITER include operating at low normalized gyroradius, maintaining a low collisionality pedestal alongside a high-density divertor, and supporting a significant population of isotropic energetic particles.

The ITER Project will significantly advance fusion energy development through an extensive research program focused on tritium breeding Test Blanket Modules, complemented by detailed physics studies and a comprehensive technology program among ITER Members.

1.3.3.2 Key Research Issues during Construction

The supporting Physics research program for ITER construction focuses on four key areas: H-mode issues, plasma-wall interaction issues, MHD instability mitigation/control issues, and scenario development issues, all of which are interconnected A primary objective of this program is to enhance integrated modeling capabilities through the Integrated Modelling and Analysis Suite (IMAS), creating a ‘flight simulator’ for experimental preparation that can be validated by current fusion facility results before application to ITER plasma discharges Additionally, the research and development activities necessary for the advancement and construction of the Test Blanket Modules are also reviewed.

Section 5 of the Research Plan outlines key research issues in construction that could significantly impact ITER These issues may affect the design of baseline components, though the opportunity for such influence has diminished due to advancements in detailed design over the past decade Additionally, they could influence the choice or design of upgradeable components and enhance the plasma scenarios and experimental plans, ultimately improving ITER's operational efficiency Time-critical issues must be resolved using existing capabilities or modest extensions, while decisions on upgrades involve complex long lead times for design, fabrication, and installation, alongside necessary R&D Lastly, challenging issues related to plasma scenario development may necessitate further upgrades, impacting the experimental schedule for ITER.

The Research Plan identifies several key issues, though it is not an exhaustive list Numerous uncertainties in the physics and technology underlying ITER could also affect its performance and are being tackled through a comprehensive research program running alongside ITER's construction The highlighted items, elaborated in section 5, are deemed the highest priority due to their potential to necessitate modifications or upgrades to the facility.

O VERALL S TRUCTURE OF THE E XPERIMENTAL P ROGRAM DURING O PERATIONS

I NTEGRATION OF THE R ESEARCH P LAN INTO THE S TAGED A PPROACH

The integration of the ITER Research Plan into the Staged Approach hinges on the connection between the ITER facility's hardware capabilities and the operational objectives at each phase A careful balance has been achieved between procurement schedule constraints and the operational needs to meet the PFPO phase goals, aiming for a transition to DT operation by late 2035 The hardware configuration evolution is detailed in section 1.3.2, which formalizes the ITER facility's baseline configuration for each Staged Approach phase Additionally, several options within this baseline are under development and will be formally adopted through the Project Change Procedure once their technical benefits and feasibility are validated Key elements of these options, which impact the Research Plan's development, are summarized in section 1.3.2, ensuring they are considered available for the experimental program.

The Research Plan outlines a logical structure for advancing from First Plasma to full performance in DT plasmas, while also estimating the duration of each step in the experimental program This provides an overall estimate of the operational time needed for successive phases, despite significant uncertainties ahead of ITER exploitation Through discussions with fusion community experts during the Research Plan Workshops, an initial estimate of the experimental time required for key activities was developed, forming the foundation for the estimates presented in subsequent subsections of section 2.

Table 2-1 provides an estimate of the maximum number of experimental days available for PFPO-1, PFPO-2, and FPO, based on previous analyses from the Research Plan and the ITER RAMI analysis [ITER_D_28WBXD, 2012] This estimate follows the operational planning outlined in the Staged Approach, which assumes a 2-shift operation consisting of 12 days dedicated to experimental activities followed by 2 days allocated for short-term maintenance within a 14-day cycle.

Table 2-1 – Overview of experimental days within the Research Plan campaigns

First Plasma Campaign (nominal 1 month) 25

Pre-Fusion Power Operation 1 (18 months) 470

Pre-Fusion Power Operation 2 (21 months) 545

Fusion Power Operation 1 (nominal 16 months) 415

Fusion Power Operation 2 (nominal 16 months) 415

Fusion Power Operation 3 (nominal 16 months) 415

In a 2-shift operational day, a 30-minute pulse repetition time permits up to 32 plasma pulses for burn durations under 450 seconds However, when accounting for the 60% 'Inherent Availability' used in the RAMI analysis and the time lost due to disruptions and recovery, the effective output is reduced to 13 'good' plasma pulses per day.

During the FPO phase, as pulses with burn durations exceeding 450 seconds are implemented, the daily pulse count will decrease while maintaining a 25% duty cycle However, this decrease in effective pulses may be offset by a lower frequency of disruptions, which is expected to be significantly reduced in the FPO phase compared to the PFPO phase This improvement is attributed to advancements in disruption management and the necessity to minimize experimental downtime caused by disruptions affecting in-vessel components at high fusion power levels.

The estimates presented in Table 2-1 represent an upper limit on the days available for the experimental program, as the system performance considered in the RAMI analysis pertains to mature systems However, these estimates serve as a valuable benchmark for comparing the anticipated experimental time needed for key activities.

Research Plan presented in the following tables

The transition from the PFPO phase, utilizing non-activating fuels, to the FPO phase, which focuses on deuterium and deuterium-tritium plasmas, is a crucial milestone in the Research Plan This shift is anticipated to occur after Integrated Commissioning IV, likely in the latter half of 2035 To support the experimental program during this transition, the Tritium Plant is set to begin commissioning its operations with tritium in spring 2033.

In 2036, during the initial phase of FPO campaign 1, the T-Plant is anticipated to provide trace amounts of tritium for the ITER experimental program.

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 Pam 3 s -1 (~ 10 23 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 to allow an expanding program of DT experiments to be developed during the latter half of the first

The FPO campaign has successfully achieved substantial fusion power production, generating several hundred megawatts This initiative establishes a foundational framework for the commissioning and operational start-up of the Tritium Plant, which is essential for advancing the FPO program detailed in section 2.6.

P LASMA -W ALL I NTERACTION (PWI) I SSUES FOR THE R ESEARCH P LAN

C HARACTERISTICS OF H ELIUM O PERATION

Recent studies have highlighted the significant impact of helium interactions with plasma-facing metals, particularly due to helium's strong trapping behavior in various metals This interaction often leads to clustering and bubble formation, resulting in notable morphological changes Furthermore, the potential development of a porous sub-surface structure, characterized by a high density of voids and bubbles, has been observed, which correlates with a marked decrease in the thermal conductivity of the surface.

Concerns have emerged regarding the potential decrease in transient damage thresholds, particularly for Edge Localized Modes (ELMs), following prolonged helium (He) plasma operations during the PFPO phases Research has demonstrated that helium-induced embrittlement occurs after high-flux plasma exposure in linear devices, indicating that exposure of tungsten (W) to He plasma can lower the threshold for surface cracking.

Figure 2.2-1: Operations Plan within the Staged Approach

ELM-simulation experiments in electron-beam facilities reveal that tungsten's cracking threshold is very low at ITER-relevant high cycle numbers (~10^6) [Wirtz, 2017] This suggests that surface cracking could occur earlier with extensive Helium operations, but it is likely to happen regardless, as Helium will be generated during DT operations.

The range of helium-induced morphological changes in tungsten varies significantly with temperature Below 600-700 K, a dense network of nanobubbles is typically observed, as illustrated in Miyamoto's 2011 study As the temperature rises to between 900-1900 K, fuzz formation occurs, highlighted in images from De Temmerman’s 2012 research At even higher temperatures, large voids are formed, further documented in De Temmerman's 2012 findings.

Research on the impact of helium (He) exposure on the mechanical properties of tungsten (W) has shown measurable increases in hardness, even at distances far from the He implantation depth However, current data remains insufficient to fully understand the long-term effects of prolonged He and plasma exposure on the thermo-mechanical properties of W, particularly concerning its power handling capabilities It is important to note that the effects of He on metals are significantly influenced by temperature During pure He operations, as anticipated in the PFPO-1 and 2 phases, the divertor heat fluxes and surface temperatures are expected to be considerably lower than those observed during high-power deuterium-tritium (DT) operations, where He is generated from fusion reactions in the divertor exhaust Consequently, insights gained from ITER regarding He-W interactions during non-active operations may not be applicable to conditions encountered in burning plasma scenarios.

In 2015, a dedicated helium (He) campaign at ASDEX-Upgrade aimed to investigate plasma-wall interactions in a fully metallic environment Samples were installed on the outer divertor target manipulator and exposed to the outer strike point The campaign revealed that the outer strike point became a deposition-dominated area during He operations, contrasting sharply with its erosion status during deuterium (D) discharges Notably, thick boron coatings were found on the samples, suggesting increased wall erosion during He operations This phenomenon is likely due to the higher average charge state and mass of He ions impacting the wall, leading to enhanced boron sputter erosion and increased ELM-induced erosion of boron coatings Consequently, the outer divertor emerged as a net boron deposition zone for He plasmas Confirming these findings in JET would be beneficial, as they imply significant material migration patterns for ITER during pure He operations.

He operations may differ significantly from the anticipated outcomes during deuterium-tritium (DT) operations, raising concerns about their effectiveness for studying beryllium co-deposition Furthermore, the distinct retention mechanisms of hydrogen isotopes compared to helium in plasma-facing materials limit the insights gained about fuel retention during helium operations.

Divertor detachment physics in helium (He) plasmas significantly differs from that in hydrogenic plasmas due to He's higher ionization energy, the presence of a second ionization stage, and the absence of molecular chemistry In He plasmas, the primary loss processes leading to detachment involve radiative loss from electron-impact excitation and ionization, whereas in hydrogen, detachment primarily occurs through momentum losses from charge-exchange processes Additionally, the longer ionization mean-free paths in He allow He neutrals to escape more easily from the target area, impacting power dynamics downstream Consequently, the distinct plasma detachment behaviors suggest that power load control techniques developed for He plasmas may not be directly applicable to high-power operations in deuterium (D) or deuterium-tritium (DT) plasmas, making H-mode plasmas a more suitable environment for developing these control techniques However, the enhanced divertor radiation in He enables operation at higher input power levels before detachment control becomes necessary, potentially addressing power load management challenges in H-mode plasmas.

S TEADY - STATE NEAR AND FAR S CRAPE -O FF L AYER (SOL) POWER WIDTHS (1/I P SCALING )

The radial profile of parallel power flow and the significance of main wall interactions in ITER remain active research topics within the fusion community, with no consensus reached yet A major concern is the predicted very narrow near-SOL inter-ELM heat flux width (λq,near) for ITER's baseline burning plasma scenario (Ip = 15 MA), which has been derived from high-resolution IR thermography measurements at the outer divertor target of various tokamaks Recent observations suggest λq,near is approximately 1.0 mm, which is about three times smaller than previously assumed in plasma boundary modeling for the ITER divertor's operational window This new scaling indicates that λq,near is inversely proportional to Ip, showing minimal dependence on other key SOL parameters identified in earlier scaling efforts.

Recent gyro-kinetic simulations at the ITER scale reveal a significant major radius dependence in the separatrix density (PSOL), with findings indicating a value of approximately 5-6 mm for q,near in 15 MA H-modes This contrasts with earlier analytic modeling based on neoclassical magnetic drift theory, which provided strong support for these concepts [Goldston, 2012].

A simulation study by Kukushkin (2013) utilizing the SOLPS plasma boundary code revealed that a very small λq,near necessitates a significant increase in upstream density to achieve the required extra dissipation in the divertor, aimed at reducing peak heat flux density This increase could exceed tolerable limits for operational density and adequate confinement Additionally, the incorporation of divertor monoblock shaping in high heat flux areas further complicates the situation, demanding even greater dissipation to keep target heat flux densities below the 10 MWm^-2 limit for steady-state power handling The research community acknowledges the critical nature of this issue for ITER and is actively working towards a consensus on expectations for λq,near at high performance levels Further plasma boundary simulations are anticipated to build on Kukushkin's findings.

2013] are also required to study the extent to which impurity seeding could be used to provide the extra dissipation in the event of very narrow heat flux channels

To accurately characterize the real q near the ITER operation, high spatial resolution infrared diagnostics will be implemented, focusing on a small toroidal area of the divertor vertical target during PFPO-1 Recent findings from ASDEX-Upgrade have contributed valuable insights to this endeavor.

Recent studies have revealed a similar parametric dependence of the heat flux width (λq,near) in L-mode discharges to that observed in H-mode, albeit with approximately double the magnitude This finding supports thermal load specifications previously adopted for ITER Additionally, new measurements indicate that λq,near is influenced by edge plasma temperature, potentially explaining the higher values seen in L-mode compared to H-mode Consequently, divertor power width measurements in ITER L-modes can inform expectations for H-mode values Furthermore, data from ASDEX-Upgrade illustrate a significant dependence of the divertor heat flux spreading parameter (S) on divertor electron temperature (Te) and density (ne), highlighting its utility in characterizing target heat loads.

The interaction of SOL power width and divertor target wetting with magnetic perturbations is crucial for effective ELM control and mitigation Current analyses, primarily conducted with the 2-D SOLPS code, overlook the impact of 3-D magnetic fields, although some studies using the 3-D EMC3-Eirene code have shown how divertor fluxes can be redirected into helical patterns at the targets These magnetic perturbations can extend significantly beyond the conventional heat flux decay profile, particularly at maximum ELM control coil currents However, existing research has mainly focused on high recycling regimes, necessitating further advancements in EMC3-Eirene to explore dissipative regimes influenced by strong seeding and recombination Additionally, understanding the effects of 3-D fields on divertor heat flux patterns in highly dissipative regimes is complicated by the fact that ELM suppression has only been observed at low collisionality Experiments indicate that in higher density divertor plasmas subjected to 3-D fields, the heat flux perturbation is typically mitigated when the plasma detaches, resembling profiles without perturbations Nevertheless, the complete suppression of ELMs may significantly alter upstream particle and heat fluxes, leading to unknown divertor target profiles under such conditions.

In a 2016 report to STAC, the ITER Organization highlighted the current status of research, emphasizing its high priority within the International Tokamak Physics Activity (ITPA) Divertor and SOL and Pedestal Topical Groups Collaborative experiments and modeling are anticipated to significantly enhance understanding in this field in the coming years.

The far-SOL power and particle fluxes on ITER, primarily influenced by filamentary or blobby cross-field convective transport, remain uncertain and are under active investigation by the ITPA Divertor and SOL Topical Group This research is crucial as it affects wall heat and particle flux magnitudes, impacting power handling, material erosion, and ICRF coupling Recent studies have established a connection between upstream density shoulders and the collisionality of divertor plasma in L-mode plasmas, while similar patterns observed in inter-ELM H-mode plasmas yield less definitive results Although divertor collisionality appears necessary for the formation of upstream density shoulders, it is not sufficient on its own Currently, there is a lack of data regarding perturbed magnetic fields for ELM control, with measurements under H-mode conditions being both challenging and infrequent.

Significant advancements have been achieved in understanding the universal characteristics of far scrape-off layer (SOL) turbulence in tokamaks, with a successful model developed to describe these statistics (Garcia, 2016) Additionally, the sophistication of SOL turbulence simulations is rapidly increasing, with the emergence of several 3-D codes that incorporate sheared (X-point) geometries Ongoing research and development, both experimental and theoretical, are expected to greatly enhance the understanding of main chamber fluxes by the time ITER operations commence.

E DGE LOCALIZED MODE (ELM) POWER WIDTH SCALING

The acceptable level of ELM energy loss (ΔWELM) concerning the erosion and damage of plasma-facing components is a dynamic issue, especially in the divertor This tolerance is influenced by the design of the monoblock front surface and the behavior of tungsten (W) material under prolonged high-temperature conditions with repeated transients Although these conditions may not reach melting points, they can still cause surface cracking, which may eventually lead to the formation of deeper macro-cracks.

The ITER Heat and Nuclear Load Specifications set a limit for the energy loss during Edge Localized Modes (ELMs), with a maximum allowed value of ΔWELM ≲ 0.6 MJ This specification was based on plasma gun experiments that demonstrated melting of aligned tungsten targets at energy densities exceeding ~0.4 MJm -2, reflecting conditions expected during uncontrolled type-I ELMs in ITER Additionally, the requirement incorporates a conservative approach that assumes no broadening on the divertor target during ELM events, alongside a specification for asymmetric energy distribution that favors the inner target.

The established criterion for power flow to an axisymmetric divertor target was initially set before addressing the specific shaping of divertor monoblocks Recent studies have explored various shaping options, considering 3-D ion Larmor orbits During edge-localized modes (ELMs), toroidal gap loading is predicted at both inner and outer targets, which surface shaping cannot fully mitigate The effects of repeated flash melting or temperatures exceeding the recrystallization threshold on these edges remain uncertain, although gradual erosion is a possibility If such erosion proves detrimental to ITER operations, divertor monoblock edge melting could impose limits on acceptable ELM sizes, potentially necessitating earlier compliance with these limits in operational campaigns than previously anticipated.

The theoretical model supporting parallel ELM energy density scaling suggests that mitigating ELMs may not significantly lower peak ELM target energy densities in ITER, as observed experimentally Instead, only full suppression could prevent ELM-induced melting at the divertor Additionally, monoblock shaping increases the stationary power flux density on the surface, and the need to stay below tungsten recrystallization temperature, particularly during ELM-driven temperature excursions, may impose stricter limits on allowable power flux densities during high-performance operation phases.

ELM interactions with the main wall have not been as thoroughly examined as those with the divertor, yet they can significantly impact erosion on ITER, particularly due to beryllium's lower melting temperature compared to tungsten The extent of ELM energy reaching the walls, influenced by ELM filament characteristics and magnetic equilibrium design, could lead to considerable melting in the secondary X-point region for large ELM energy losses While this issue is not anticipated until the burning plasma phase, detailed calculations for gap loading on the first wall have yet to be conducted Any melting of beryllium tiles is expected to pose less of a problem, as the plasma is more tolerant of beryllium impurities, and the management of power fluxes on the blanket first wall is less critical than in the divertor.

A MMONIA FORMATION DURING N 2 SEEDING

Ammonia (NH3) formation occurs in tokamaks when nitrogen is introduced into the scrape-off layer (SOL) and divertor plasma to mitigate heat fluxes Research conducted in the all-metal JET and ASDEX-Upgrade tokamaks indicates that up to 8% of the injected nitrogen can be transformed into ammonia Tritium-laden ammonia is effectively retained by the active charcoal in ITER's cryopumps, necessitating complete regeneration of the cryopumps for recovery The tritium inventory in these cryopumps is limited to 180 g, and regular regenerations at approximately 100 K will be conducted during operations to retrieve the pumped hydrogenic species.

Regenerating ammonia at elevated temperatures is feasible but time-intensive, taking approximately six hours per pump, with six pumps in ITER This lengthy process could significantly disrupt the ITER duty cycle due to substantial ammonia production Additionally, the creation of tritiated ammonia influences the design of the tritium plant, which must effectively decompose ammonia to recover tritium while preventing the formation of harmful nitrogen oxides.

Ammonia formation occurs through surface reactions between nitrogen and hydrogenic radicals, but it remains uncertain whether these reactions take place in plasma-exposed or plasma-shadowed areas As a result, predicting ammonia formation rates in ITER is currently not feasible Ongoing experimental activities in tokamaks and laboratory devices aim to clarify the ammonia formation process under fusion-relevant conditions and to quantify the reaction rates of the involved elementary steps Continued efforts in this research are essential for integrating the relevant nitrogen chemistry into fluid plasma-boundary codes.

D IFFERENCES IN MEDIUM -Z IMPURITY SEEDING (N 2 VS N E )

Recent advancements in major tokamak experiments featuring all-metal walls, particularly JET and ASDEX-Upgrade, highlight the challenges of replicating plasma performance in highly radiative scenarios When extrinsic seeding is utilized to substitute natural carbon impurities for divertor power flux control, achieving similar results as those observed with all-carbon plasma-facing components (PFCs) proves to be difficult.

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

The formation of tritiated ammonia during nitrogen injection poses a significant challenge for ITER, primarily limiting the achievable duty cycle In contrast, neon emerges as a more favorable option, as it is a fully recycling gas that effectively prolongs the timescales for preventing divertor monoblock overheating during loss of seeding or other divertor reattachment events However, recent experiments at JET have shown that when wall surfaces become saturated with nitrogen, the timescales for losing detachment can be considerably extended if the seeding gas is withdrawn.

ITER fundamentally differs from machines like ASDEX-Upgrade and JET due to its size, magnetic field strength, and the ability to achieve high pedestal temperatures Current experiments indicate that in nitrogen-seeded plasmas, radiation primarily occurs in the divertor, whereas with neon, main chamber radiation increases without a corresponding rise in divertor radiation This increased main chamber radiation reduces power crossing the separatrix and influences ELM frequencies Moreover, the introduction of impurities alters pedestal transport; nitrogen enhances pedestal top temperatures and confinement, while neon injection elevates pedestal top density in ASDEX-Upgrade but significantly decreases pedestal pressure in JET Both devices experience instability when neon injection exceeds a certain threshold.

In ASDEX-Upgrade, the reduction in ELM frequency leads to decreased ELM impurity flushing, while the increased pedestal density gradient intensifies inward transport of Ne and W, resulting in heightened core radiation losses These factors, combined with Ne's inability to diminish divertor power loads through increased SOL/divertor radiation, complicate the study of power exhaust with Ne, despite ASDEX-Upgrade operating at P sep /R values akin to those anticipated in ITER's burning plasmas Conversely, in JET, where Psep/R values are approximately half those of ASDEX-Upgrade, Ne reduces power flow across the separatrix, causing H-L back transitions, which complicates the comparison between nitrogen and neon until additional heating power becomes available.

Radiative divertor experiments on Alcator C-Mod with molybdenum plasma-facing components (PFCs) at high magnetic fields demonstrated that both nitrogen (N2) and neon (Ne) seeding in EDA modes achieved partially detached divertor conditions at a normalized confinement of H98 ~1 This occurred with edge power flows only slightly exceeding the H-mode threshold power by factors of 1 to 1.4, similar to conditions in ITER As observed in JET and ASDEX-Upgrade experiments, high radiative fractions primarily resulted in radiation occurring in the X-point region, with some extension onto flux surfaces just inside the separatrix However, in C-Mod, this radiation was insufficient to significantly impact the pedestal top, even in cases involving Ne.

Current challenges hinder the confident extrapolation of findings from existing devices to ITER, particularly regarding performance with low-Z impurity seeding Simulations using SOLPS for burning plasma conditions indicate that ITER will compress both the power flow and the plasma behavior under equivalent SOL power conditions.

Ne and N are strongly present in the divertor, particularly in the strike point regions, with Ne radiation being distributed over a somewhat larger volume Although similar partially detached conditions can be achieved, different seeding rates are necessary for each case Integrated modeling, utilizing the boundary plasma scaling approach, has demonstrated these findings.

N2 and Ne offer sufficient operational windows for burning plasma, with N2 demonstrating a slight advantage due to its lower core radiation compared to Ne However, further research is needed to enhance the integrated model and broaden the scope of N2 seeded boundary cases, as Ne seeding has historically been the preferred method for power flux control at ITER.

The substantial physical size of ITER, combined with the anticipated higher pedestal and separatrix temperatures, suggests that the current differences observed between nitrogen (N2) and neon (Ne) will diminish at the ITER scale Elevated pedestal and separatrix temperatures are expected to result in reduced Ne radiation within the core plasma Additionally, in the scrape-off layer (SOL) and divertor, the mean-free paths for ionization of neutral Ne surpass those of nitrogen, leading to varying influences of thermal and friction forces on the two species Consequently, the net parallel force in the divertor is lower for Ne compared to N2.

N Both effects drive lower divertor impurity enrichment for Ne, but the differences are likely to be much reduced with increasing device size

Research on low-Z impurity seeding is progressing rapidly, with both experimental and theoretical advancements anticipated in the near future A key focus will be on integrating ELM control with extrinsic seeding to manage power flux effectively at high performance levels Upcoming experiments on JET, utilizing maximum input power and current to achieve pedestal temperatures similar to those expected in ITER, aim to bridge the operational gap to ITER These experiments will compare the efficiency of N2 and Ne seeding for divertor power flux control while maintaining acceptable core plasma performance.

T UNGSTEN - RELATED MATERIAL ISSUES

Tungsten is often praised for its high melting point as a plasma-facing material; however, recrystallization and grain growth occur at significantly lower temperatures, leading to notable changes in its mechanical properties This process results in decreased strength, hardness, and shock resistance, despite an increase in ductility During high heat flux testing at 20 MW/m², the emergence of macro-cracks in tungsten is linked to the loss of ductility in recrystallized tungsten, as the testing temperature is sufficient to trigger recrystallization.

To mitigate the negative impacts of recrystallization, it is essential to establish an operational budget that considers the time and temperature factors of the recrystallization process Currently, there is limited understanding of the recrystallization kinetics specific to the tungsten intended for the ITER divertor, necessitating further research to refine the operational budget effectively.

Even when bulk recrystallization is minimized, electron laser melting (ELM) can still lead to significant surface damage, including roughening, cracking, and melting, which are influenced by the energy density of the ELM Extensive research has been conducted on the evolution of this surface damage, focusing on factors such as base material temperature, energy density, and pulse number, particularly in electron-beam facilities and using advanced pulsed plasma or laser systems Most studies have concentrated on pristine materials without the influence of particle irradiation.

Research indicates that combined particle and heat loading can create synergetic effects, leading to a lower damage threshold on plasma-exposed surfaces A significant uncertainty for ITER lies in understanding how these effects might influence material compatibility with high-performance plasmas Currently, it is challenging to establish an acceptable damage level before material properties begin to deteriorate, potentially affecting plasma performance through increased tungsten release or diminished power handling capabilities This highlights the necessity for further investigation prior to the commencement of ITER operations.

The long-term effects of stationary and transient plasma exposure, whether from hydrogen, deuterium, or helium, on the thermo-mechanical properties of tungsten (W) remain underexplored, despite indications that plasma exposure alters material characteristics Current extrapolations of these findings to the high fluence levels anticipated in ITER are fraught with uncertainty To accurately evaluate the implications for ITER, it is essential to incorporate these effects into the modeling of the material's thermo-mechanical response.

2017] for example) and in turn would require knowledge of the evolution of material properties with plasma fluence (ultimate strength, toughness, thermal conductivity, etc.).

C ASTELLATION GAP HEAT LOADS

Since the last major update of the IRP, significant advancements have been achieved in divertor design, notably the removal of the initial carbon/tungsten divertor from ITER's operational plans in 2011 This pivotal decision prompted an expedited design process for the full tungsten version, which is now essential for PFPO-1 and must ensure reliable operation throughout the FPO phases, including considerations for lifetime requirements and potential divertor upgrades.

One important design area that had not been addressed for the full-W divertor was the question of

The high stationary power flux densities achievable by ITER pose a significant challenge for the durability of metal divertor targets, particularly concerning melting at the leading edges of gaps between neighboring monoblocks due to manufacturing and assembly tolerances Extensive design and physics studies have led to a finalized shaping of plasma-facing components (PFCs) in both high heat flux and baffle regions of vertical targets These studies reveal the intricate nature of plasma interactions near PFC gaps, as plasma ions can access gap edges in both poloidal and toroidal directions, especially during ELM transients, despite being magnetically shadowed in a simplified optical model.

The intricate monoblock surface shaping used in the JET bulk W outer divertor target is impractical for the ITER divertor design, which instead employs a simple toroidal bevel to mitigate misalignment issues in high heat flux areas However, this design does not address the loading of gaps between neighboring blocks in the toroidal direction Shaping the components significantly decreases the allowable parallel power flux density, with the current ITER baseline toroidal bevel reducing this by approximately 40% compared to an unshaped solution, thereby limiting the stored energy loss per ELM to prevent toroidal gap edge melting While a poloidal bevel can protect toroidal gaps, it is only effective at the outer target and further restricts the allowable stationary parallel heat flux densities.

The ITER divertor's numerous gap edges necessitate meticulous attention to power loading as operational performance enhances Gap edge loading, occurring at the Larmor radius scale of a few hundred microns, requires high-resolution divertor infrared systems to effectively detect interactions in small regions of the vertical targets In contrast, lower resolution machine protection infrared systems can monitor stationary top surface loading across many individual monoblocks in strike point areas Recent research and development prompted by the ITER divertor design requirements should persist, particularly in verifying ELM-driven ion orbit loading Additionally, ongoing theoretical studies of ELM evolution and transport are crucial for refining specifications related to particle energies and time envelopes essential for the ITER monoblock shaping.

R ADIO -F REQUENCY - ASSISTED WALL CONDITIONING TECHNIQUES

The continuous maintenance of the toroidal magnetic field by superconducting magnets during ITER operational periods significantly decreases the efficiency of DC glow discharges To address this, Ion Cyclotron Wall Conditioning (ICWC) is planned for use between plasma pulses, allowing for effective fuel and impurity removal while operating within the toroidal field This approach will also help mitigate the accumulation of tritium inventory in the machine during the First Plasma Operation (FPO) campaigns.

In the absence of the ICRF system in PFPO-1, Electron Cyclotron Wall Conditioning (ECWC), which is compatible with Bt, can be utilized Both ECWC and ICRF employ conventional heating antennas to generate low-temperature magnetized discharges In ECWC, the coupling of power to the discharge primarily occurs through the collisional absorption of radio-frequency (RF) energy by electrons, allowing discharges to be produced at any Bt However, ECWC discharges exhibit poloidal and radial non-uniformities due to RF wave absorption Improvements in discharge homogeneity and wall coverage have been achieved through monopole phasing and mode conversion schemes For ITER, ECWC is anticipated to require 1–5 MW of power with a duty cycle of 1 second on and 30 seconds off to create a He, H, or D plasma with electron density around 10^17.

ECWC discharges are generated only within the Electron Cyclotron Resonance (ECR) layer inside the vacuum vessel, where high plasma densities (ne ∼ 10^18 - 10^19 m^-3) can impact discharge uniformity and cleaning efficiency, similar to ICWC but to a greater extent Although pulsed operation helps mitigate these effects, low absorption per pass in the optically thin ECWC plasma poses risks of excessive stray radiation exposure to plasma-facing components (PFCs) and diagnostics The operational parameters for ECWC, including pressure and power, are expected to closely align with those of the ECRH-assisted breakdown planned for ITER, potentially requiring 1–10 MW at 170 GHz with a duty cycle akin to ICWC.

In both techniques utilizing hydrogen (H) or deuterium (D) as working gases, the primary flux to the wall consists of isotropic neutral hydrogenic species These species either desorb from wall surfaces or arise from molecular dissociation The ion flux parallel to the magnetic field lines peaks at limiting surfaces, typically reaching around 10²¹ ions m⁻² s⁻¹, and diminishes exponentially towards the plasma edge, decreasing to a few 10¹⁹ ions m⁻² s⁻¹.

In comparison to GDC plasma, which has ion fluxes ranging from 10^17 ions·m^-2·s^-1, ICWC generates fast charge exchange neutrals with temperatures exceeding 1 keV and energies reaching up to 50 keV The significant penetration depths of these particles, capable of reaching up to 200 nm in beryllium, enhance their effectiveness for T-removal applications.

ICWC is routinely applied between shots in the KSTAR tokamak for impurity removal [Kwon,

In 2011, studies indicated that the Inter-Coolant Water Cooling (ICWC) system could effectively reduce tritium (T) inventory build-up in ITER, with extrapolations from current devices, particularly experiments conducted at JET with ITER-like walls operating at 200°C, suggesting that ICWC could remove approximately 0.35 to 0.6 grams of tritium between plasma shots This removal rate is comparable to the estimated retention per pulse, highlighting the potential of ICWC in managing tritium levels efficiently.

Recent studies indicate that surface temperature significantly influences isotopic exchange efficiency Consequently, conducting new experiments at the JET facility with reduced wall temperatures, specifically operating the ICWC at 70°C as planned for ITER, is crucial for understanding this relationship.

The efficiency of Edge Cooling Water Cooling (ECWC) systems utilizing metallic plasma-facing components (PFCs) is not extensively documented Successful recovery from disruptions using ECWC has only been demonstrated on the JT60-U device Additionally, limited published data on fuel removal in devices with carbon PFCs suggests that their efficiency is lower compared to that of Integrated Cooling Water Cooling (ICWC) systems.

D IVERTOR DETACHMENT CONTROL

Effective control of divertor power loading is crucial for ITER's operation, requiring a robust and reliable system The W divertor monoblock technology, designed for vertical targets, can handle steady-state surface loads of approximately 10 MW/m² However, to prevent recrystallization, exposure to higher loads must be limited, and severe excursions that cause global surface melting must be avoided to prevent significant topological deformation Sustained critical heat flux at the cooling interface can lead to catastrophic failures, including water leaks and the need for extensive in-vessel repairs While such heat flux densities typically occur during high-current burning plasma operations, heat flux control may also be necessary during PFPO-2 operations, where SOL input powers could exceed 40 MW It is essential to develop and implement detachment control during pre-nuclear operational phases prior to the introduction of deuterium and tritium.

Power flux control in current devices has seen significant advancements recently, particularly with the implementation of tungsten (W) divertor targets that are inertially cooled, as observed in ASDEX-Upgrade and JET This technology is essential for enabling high power operations, even for short pulse durations of seconds The emergence of several medium-sized superconducting divertor devices, such as EAST and WEST with actively cooled W divertors, and KSTAR's plans to upgrade to full-W actively cooled components by 2021, indicates that progress in managing heat loads is likely to become more prevalent in the field.

The ASDEX-Upgrade has produced the most advanced detachment controller to date, utilizing current measurements from an outer target shunt This system incorporates real-time processing to eliminate ELM spikes, allowing for accurate assessment of the thermoelectric current flowing into the outer divertor, which is influenced by the inner target.

The electron temperature (Te) between edge localized modes (ELMs) serves as a reliable approximation for the outer target Te, providing a measure of the detachment state This signal is utilized within a proportional-integral (PI) controller to adjust the impurity gas puffing rate needed to achieve a specific divertor Te The ASDEX-Upgrade divertor effectively employs nitrogen as a direct actuator for divertor radiation due to its efficient compression Recently, a real-time double radiative feedback technique has been implemented at ASDEX-Upgrade to control the simultaneous injection of argon (Ar) in the main chamber and nitrogen in the divertor This system leverages a combination of foil bolometer channels in both the main chamber and the outer divertor leg Other tokamaks have explored feedback mechanisms using bolometer chords, while a real-time nitrogen feedback control technique was also tested in JET, utilizing ELM-filtered outer target Langmuir probe ion current densities to assess detachment levels An earlier effort on JET involved seeding control through a VUV nitrogen line.

Due to the intricate design of the water-cooled monoblock plasma-facing units, divertor shunts cannot be integrated into the ITER divertor targets, necessitating the development of power flux control methods based on available sensors ITER will feature a comprehensive array of bolometer channels and IR thermography, covering approximately 90% of the high heat flux areas of the divertor target Although the spatial resolution may be coarse in many regions, the toroidally symmetric nature of divertor target loading allows for the selection of zones with higher spatial resolution However, IR surface power measurements have not yet been utilized for heat flux control in tokamaks, leaving a gap in the feedback mechanisms for the critical parameter of surface heat flux density Challenges in measuring power fluxes in detaching/recombining divertor plasmas, such as intense bremsstrahlung radiation and metallic reflections, complicate the reliability of IR cameras operating in the 3-5 μm band Historically, research has leaned towards more straightforward and indirect techniques, like feedback from bolometer channels Nevertheless, the extensive IR coverage planned for ITER underscores the need for intensified R&D efforts to demonstrate real-time IR-based heat flux control for metallic components in a highly dissipative regime Initial development of a real-time capable IR system has commenced at ASDEX-Upgrade and is anticipated to be tested soon.

ITER will be equipped with additional sensors, including arrays of divertor Langmuir probes and thermocouples, to enhance detachment control loops The divertor neutral pressure is crucial for assessing detachment levels, and ASDEX-type fast neutral pressure gauges will be installed in various ITER divertor cassettes for potential use in control loops However, the longevity of in-vessel diagnostics is a significant concern, as extensive electrical cabling may not endure the harsh conditions during the nuclear phases when robust heat flux control is essential.

PWI/ PLASMA BOUNDARY CODE VALIDATION

As ITER operations approach, it is crucial for the community's modeling tools focused on plasma-wall interactions and plasma boundary physics to achieve a high level of maturity and validation through experimental benchmarks While 2-D edge plasma fluid codes have largely matured, challenges persist in accurately replicating electric fields and plasma flows in the scrape-off layer (SOL), as well as incorporating drifts at ITER scales These issues are often linked to the implicit Maxwellian fluid assumptions in these models, suggesting that a more accurate representation of plasma behavior may require the integration of kinetic effects However, purely kinetic approaches demand substantial computational resources, making them challenging to implement at the scale necessary for ITER.

As computing speeds are anticipated to improve, advancements in computer performance often balance out with refinements in physics models, maintaining consistent run-times During ITER operations, there is a critical demand for a fast and reliable plasma edge solver integrated with an appropriate core model for online analysis and discharge preparation Fully consistent core-edge coupling models are just beginning to be developed The use of magnetic perturbations for ELM mitigation necessitates a 3-D representation of the ITER boundary plasma Current tools face challenges in addressing the highly dissipative regime of the partially detached ITER divertor plasma, requiring significant efforts to overcome these obstacles Additionally, conventional transport codes must incorporate turbulence approaches to provide a realistic description of blobby intermittent transport, which significantly influences particle and heat loads on the main chamber walls Accurate modeling of the fluxes impacting the walls, through comprehensive 2-D or 3-D models, will enhance the understanding of wall material migration, redeposition patterns, and fuel retention in plasma-facing components.

Significant advancements in material migration modeling have been made to enhance the understanding of erosion and deposition effects, as well as to refine fuel retention estimates Researchers have adopted two primary approaches: one focusing on local effects at the component or regional level, and the other on global poloidally-resolved migration The local approach incorporates the 3-D geometry of ITER's first wall panels, allowing for an analysis of how component shaping influences material redistribution Conversely, the global approach effectively replicates retention rates observed in JET and the average erosion/deposition patterns When applied to ITER, this modeling indicates that tritium co-deposition with beryllium may primarily occur in either the divertor or the first wall, influenced by the far-SOL plasma profiles Given the intricately shaped ITER first wall, three-dimensional effects are crucial, highlighting the need for advanced global 3-D material migration models These models will enhance the understanding of erosion and deposition processes in ITER and are closely linked to the development of 3-D boundary SOL codes, which provide essential plasma background information for studying impurity generation and transport.

Migration modeling focuses on understanding the deposition of eroded particles within the vessel and quantifying the tritium trapped in these deposited layers, especially for ITER, which features a low-Z first wall with a low erosion energy threshold for ion impacts Previous studies have utilized one-dimensional diffusion/trapping models to evaluate the efficiency of tritium removal from co-deposits during plasma-facing component (PFC) bake-out However, this 1-D approach fails to provide a comprehensive model for the outgassing process, as different regions of the machine are heated to varying temperatures—240 °C for the first wall and 350 °C for the divertor Additionally, tritium retention in co-deposits is highly spatially dependent on local co-deposition conditions A potential solution to these challenges lies in coupling diffusion/trapping models with migration codes.

Modeling plasma-material interactions poses significant challenges due to the vast range of temporal and spatial scales involved, from atomic-level ion/surface interactions to the effects on the mechanical properties of plasma-facing materials Developing multi-scale models to predict the evolution of thermal and mechanical properties over time is essential, yet currently unattainable While some progress has been made in understanding hydrogen and helium interactions with tungsten surfaces, further expansion of these efforts is necessary Concurrent advancements in experimental techniques, such as micro-mechanical measurements and in-situ diagnostic systems, will provide crucial data for validating these models To create reliable workflows that offer a coherent understanding of ITER plasma and its interaction with surrounding components, it is vital to integrate these activities within the ITER Integrated Modelling and Analysis Suite (IMAS).

F IRST P LASMA

P RE -F USION P OWER O PERATION P HASE (PFPO)

MA/2.65 T

F USION P OWER O PERATION P HASE (FPO)

T EST B LANKET M ODULE (TBM) T ESTING P ROGRAM IN ITER

H EATING AND CURRENT DRIVE SYSTEMS

U PGRADES OF PLASMA - FACING COMPONENTS AND MATERIALS

U PGRADES OF DISRUPTION MITIGATION SYSTEMS

H- MODE ISSUES

MHD STABILITY ISSUES

S CENARIO DEVELOPMENT ISSUES