Steady-state Operation In ITER Research Articles

Microstructural changes induced in advanced tungsten grades under high temperature neutron irradiation

Recent studies addressing the change of mechanical properties after high temperature irradiation of tungsten have reported substantial increase of the ductile to brittle transition temperature as well as increase of the hardness. In this work, we performed microstructural analysis by means of transmission electron microscopy (TEM) on various tungsten grades earlier exposed to high temperature neutron irradiation at 1200 °C up to 1 dpa. The latter corresponds to the expected irradiation temperature on the surface of a tungsten monoblock during the steady state operation in ITER. TEM was applied to five tungsten (W) materials including ITER specification pure W and precipitate strengthened W alloys. The analysis of TEM data coupled with its discussion led to a number of important conclusions regarding the damage induced at 1200 °C neutron irradiation such as: (i) weak effect of recrystallization on the accumulation of the lattice damage; (ii) promising results on the tolerance of damage accumulation in W-Y2O3 grade and (iii) internal oxidation of TiC particles which may contribute to the huge increase of the irradiation induced hardening.

Three-dimensional dynamics of fluctuations appearing during pellet ablation process around a pellet in a fusion plasma experiment

Understanding pellet ablation physics is crucial to realizing efficient fueling into a high temperature plasma for the steady state operation of ITER and future fusion reactors. Here we report the first observation of the formation of fluctuation structures in the pellet plasmoid during the pellet ablation process by a fast camera in a medium-sized fusion device, Heliotron J. The fluctuation has a normalized fluctuation level of ~ 15% and propagates around the moving pellet across the magnetic field. By comparing the fluctuation structures with the shape of magnetic field lines calculated with the field line tracing code, we successfully reconstruct the spatio-temporal structure of the fluctuations during the pellet ablation process. The fluctuations are located at the locations displaced toroidally from the pellet and propagate in the cross-field direction around the pellet axis along the field line, indicating a three-dimensional behavior and structure of fluctuations. The fluctuation would be driven by a strong inhomogeneity formed around the pellet and invoke the relaxation of the gradient through a cross-field transport induced by the fluctuations, which could affect the pellet ablation and pellet fueling processes. Such fluctuations can be ubiquitously present at the inhomogeneity formed around a pellet in the pellet ablation process in fusion devices.

Integrated control of individual plasma scalars with simultaneous neoclassical tearing-mode suppression

A novel integrated-control architecture has been tested in nonlinear, one-dimensional simulations using the control-oriented transport simulator (COTSIM©) and in DIII-D experiments. Integrated architectures that can perform continuous-mission control while also handling off-normal events will be vital in future reactor-grade tokamaks. Continuous-mission controllers for individual magnetic and kinetic scalars (thermal stored-energy (W), volume-average toroidal rotation (Ω ϕ ), and safety factor profile (q) at different spatial locations) have been integrated in this work with event-triggered neoclassical tearing-mode (NTM) suppression controllers by combining them into an architecture augmented by a supervisory and exception handling (S&EH) system and an actuator management (AM) system. The AM system, which enables the integration of competing controllers, solves in real time a nonlinear optimization problem that takes into account the high-level control priorities dictated by the S&EH system. The resulting architecture offers a high level of integration and some of the functionalities that will be required to fulfill the advanced-control requirements anticipated for ITER. Initial simulations using COTSIM suggest that the plasma performance and its MHD stability may be improved under integrated feedback control. In addition, the integrated-control architecture has been implemented in the DIII-D plasma control system and tested experimentally for the first time ever in DIII-D in a high-q min scenario, which is a candidate for steady-state operation in ITER.

Neutron irradiation hardening across ITER diverter tungsten armor

In this work, we have performed neutron irradiation and sub-sequent hardness measurements on a series of tungsten grades to screen the irradiation-induced hardness as a function of irradiation temperature reaching up to 1200 °C. The selected irradiation temperatures were chosen by performing temperature analysis of the expected irradiation temperature on tungsten monoblock during the steady state operation in ITER, where 1200 °C corresponds to the surface temperature at 10 MW/m2 flux density expected during normal operational conditions. The applied neutron fluence and flux (using BR2 material test reactor, up to 1 dpa) is representative of ITER irradiation conditions except the neutron spectrum. However, the measures were taken to reduce the thermal neutron flux to limit the transmutation closer to the fusion conditions. The irradiation-induced hardness measured in single crystal after irradiation at 600–800 °C agrees very well with the earlier data reported after HFIR irradiation experiments. The new irradiation data obtained in the temperature range 900–1200 °C show that even at one third melting point the neutron exposure raises the hardness by 40% to 70%, depending on the selected grade. Screening measurements by transmission electron microscopy, applied to clarify the origin of the hardening at 1200 °C, have proven the presence of the dislocation loops and high density of voids. The presence of those defects should imply the reduction of thermal conductivity, fracture toughness as well as alteration of hydrogen isotope permeation and trapping.

Reassessment of steady-state operation in ITER with NBI and EC heating and current drive

One of the three goals of the ITER project is to demonstrate fusion gain Q ≥ 5 in steady-state operation (SSO). A reassessment of the necessary conditions for Q ≥ 5 SSO in ITER has been carried out for steady-state scenarios assuming no internal transport barrier in the core plasma and with the sole use of neutral beam injection (NBI) and electron cyclotron heating and current drive (EC H&CD) as heating and current drive sources in these scenarios. The parametric operational space for SSO in ITER has been reassessed utilizing an inverse evaluation approach that takes into account the baseline design of the NBI and EC H&CD systems with PNBI = 33 MW, PEC = 20 MW and their possible upgrade capabilities included in their design, PNBI ≤ 49.5 MW, and PEC ≤ 30 MW. The optimal operational points from this evaluation approach have been chosen for detailed 1.5-D transport modelling by ASTRA and followed-up by magnetohydrodynamic (MHD) stability analysis to demonstrate the conditions in which the Q = 5 SS goal can be achieved in ITER where the plasma current is fully accounted for by the driven current by NBI and EC H&CD and the self-driven bootstrap current. The ideal MHD stability of plasma configurations with self-consistently simulated plasma profiles and equilibrium for the chosen OPs has been analysed by the KINX code. Using this analysis, the possibility to control the MHD stability of these steady-state plasmas by tailoring the current profile with the flexibility provided by design of the systems for electron cyclotron current drive and neutral beam current drive in ITER has been demonstrated. Issues related to the realization of such scenarios from the point of view of plasma physics, experimental demonstration and design limits of the ITER systems and components is discussed.

Progress of physics understanding for long pulse high-performance plasmas on EAST towards the steady-state operation of ITER and CFETR

Recently, the first ever 100 s long, steady-state H-mode discharge with good control of impurities, core and edge MHD stabilities, and heat exhaust was demonstrated in the Experimental Advanced Superconducting Tokamak (EAST) using the ITER-like (International Tokamak Experimental Reactor) tungsten upper divertor. Using both radio frequency (RF) power and neutral beam injection (NBI) heating, EAST has demonstrated fully non-inductive scenarios with an extension of fusion performance at high density and low rotation: β P ∼ 2.5, β N ∼ 2.0, H98,y2 ∼ 1.2, bootstrap current fraction fBS ∼50% at q95 ∼ 6.8. With pure RF power heating, plasmas have been maintained for up to 21 s (over 40 times the current relaxation time) with zero loop voltage and small edge localized modes (ELMs) at high density (ne/nGW ∼ 0.6–0.8), β P ∼ 2.0, β N ∼ 1.6, and ƒBS ∼47%. Experimental investigations show how plasma current profiles, turbulent transport and radiation properties self-consistently evolve toward fusion relevant steady state conditions. Modeling and physics experiments have confirmed the synergistic effects between electron cyclotron heating (ECH) and low hybrid wave (LHW), where ECH enhances the heating and current drive from LHW injection, enabling fully non-inductive operation at higher density. Small/no ELMs facilitate the RF power coupling in the H-mode phase and reduce divertor erosion. A low tungsten concentration was observed at high β P with a hollow profile in the core. Reduction of the peak divertor heat flux with f rad of up to 40% was compatible with the high β P scenario by using active radiation feedback control. With features such as dominant electron heating, zero/low NBI torque and an ITER-like tungsten divertor, fully non-inductive high-performance experiments on EAST offer unique contributions towards the succesful operation of ITER and CFETR (the Chinese Fusion Engineering Testing Reactor).

Realization of minute-long steady-state H-mode discharges on EAST

In the 2016 EAST experimental campaign, a steady-state long-pulse H-mode discharge with an ITER-like tungsten divertor lasting longer than one minute has been obtained using only RF heating and current drive, through an integrated control of the wall conditioning, plasma configuration, divertor heat flux, particle exhaust, impurity management, and effective coupling of multiple RF heating and current drive sources at high injected power. The plasma current (Ip ∼ 0.45 MA) was fully-noninductively driven (Vloop < 0.0 V) by a combination of ∼2.5 MW LHW, ∼0.4 MW ECH and ∼0.8 MW ICRF. This result demonstrates the progress of physics and technology studies on EAST, and will benefit the physics basis for steady state operation of ITER and CFETR.

High internal inductance for steady-state operation in ITER and a reactor

Increased confinement and ideal stability limits at relatively high values of the internal inductance () have enabled an attractive scenario for steady-state tokamak operation to be demonstrated in DIII-D. Normalized plasma pressure in the range appropriate for a reactor has been achieved in high elongation and triangularity double-null divertor discharges with at , near the ideal kink stability limit calculated without the effect of a stabilizing vacuum vessel wall, with the ideal-wall limit still higher at . Confinement is above the H-mode level with . At , the current is overdriven, with bootstrap current fraction , noninductive current fraction and negative surface voltage. For ITER (which has a single-null divertor shape), operation at is a promising option with and the remaining current driven externally near the axis where the electron cyclotron current drive efficiency is high. This scenario has been tested in the ITER shape in DIII-D at , so far reaching and at with performance appropriate for the ITER Q=5 mission, . Modeling studies explored how increased current drive power for DIII-D could be applied to maintain a stationary, fully noninductive high discharge. Stable solutions in the double-null shape are found without the vacuum vessel wall at , and , and at with the vacuum vessel wall.

The high-βN hybrid scenario for ITER and FNSF steady-state missionsa)

New experiments on DIII-D have demonstrated the steady-state potential of the hybrid scenario, with 1 MA of plasma current driven fully non-inductively and βN up to 3.7 sustained for ∼3 s (∼1.5 current diffusion time, τR, in DIII-D), providing the basis for an attractive option for steady-state operation in ITER and FNSF. Excellent confinement is achieved (H98y2 ∼ 1.6) without performance limiting tearing modes. The hybrid regime overcomes the need for off-axis current drive efficiency, taking advantage of poloidal magnetic flux pumping that is believed to be the result of a saturated 3/2 tearing mode. This allows for efficient current drive close to the axis, without deleterious sawtooth instabilities. In these experiments, the edge surface loop voltage is driven down to zero for >1 τR when the poloidal β is increased above 1.9 at a plasma current of 1.0 MA and the ECH power is increased to 3.2 MW. Stationary operation of hybrid plasmas with all on-axis current drive is sustained at pressures slightly above the ideal no-wall limit, while the calculated ideal with-wall MHD limit is βN ∼ 4–4.5. Off-axis Neutral Beam Injection (NBI) power has been used to broaden the pressure and current profiles in this scenario, seeking to take advantage of higher predicted kink stability limits and lower values of the tearing stability index Δ′, as calculated by the DCON and PEST3 codes. Results based on measured profiles predict ideal limits at βN > 4.5, 10% higher than the cases with on-axis NBI. A 0-D model, based on the present confinement, βN and shape values of the DIII-D hybrid scenario, shows that these plasmas are consistent with the ITER 9 MA, Q = 5 mission and the FNSF 6.7 MA scenario with Q = 3.5. With collisionality and edge safety factor values comparable to those envisioned for ITER and FNSF, the high-βN hybrid represents an attractive high performance option for the steady-state missions of these devices.

External heating and current drive source requirements towards steady-state operation in ITER

Steady state scenarios envisaged for ITER aim at optimizing the bootstrap current, while maintaining sufficient confinement and stability to provide the necessary fusion yield. Non-inductive scenarios will need to operate with internal transport barriers (ITBs) in order to reach adequate fusion gain at typical currents of 9 MA. However, the large pressure gradients associated with ITBs in regions of weak or negative magnetic shear can be conducive to ideal MHD instabilities, reducing the no-wall limit. The E × B flow shear from toroidal plasma rotation is expected to be low in ITER, with a major role in the ITB dynamics being played by magnetic geometry. Combinations of heating and current drive (H/CD) sources that sustain reversed magnetic shear profiles throughout the discharge are the focus of this work. Time-dependent transport simulations indicate that a combination of electron cyclotron (EC) and lower hybrid (LH) waves is a promising route towards steady state operation in ITER. The LH forms and sustains expanded barriers and the EC deposition at mid-radius freezes the bootstrap current profile stabilizing the barrier and leading to confinement levels 50% higher than typical H-mode energy confinement times. Using LH spectra with spectrum centred on parallel refractive index of 1.75–1.85, the performance of these plasma scenarios is close to the ITER target of 9 MA non-inductive current, global confinement gain H98 = 1.6 and fusion gain Q = 5.

Physics of the conceptual design of the ITER plasma control system

The ITER plasma control system (PCS) will play a central role in enabling the experimental program to attempt to sustain DT plasmas with Q=10 for several hundred seconds and also support research toward the development of steady-state operation in ITER. The PCS is now in the final phase of its conceptual design. The PCS relies on about 45 diagnostic systems to assess real-time plasma conditions and about 20 actuator systems for overall control of ITER plasmas. It will integrate algorithms required for active control of a wide range of plasma parameters with sophisticated event forecasting and handling functions, which will enable appropriate transitions to be implemented, in real-time, in response to plasma evolution or actuator constraints.In specifying the PCS conceptual design, it is essential to define requirements related to all phases of plasma operation, ranging from early (non-active) H/He plasmas through high fusion gain inductive plasmas to fully non-inductive steady-state operation, to ensure that the PCS control functionality and architecture will be capable of satisfying the demands of the ITER research plan. The scope of the control functionality required of the PCS includes plasma equilibrium and density control commonly utilized in existing experiments, control of the plasma heat exhaust, control of a range of MHD instabilities (including mitigation of disruptions), and aspects such as control of the non-inductive current and the current profile required to maintain stable plasmas in steady-state scenarios. Control areas are often strongly coupled and the integrated control of the plasma to reach and sustain high plasma performance must apply multiple control functions simultaneously with a limited number of actuators. A sophisticated shared actuator management system is being designed to prioritize the goals that need to be controlled or weigh the algorithms and actuators in real-time according to dynamic control needs. The underlying architecture will be event-based so that many possible plasma or plant system events or faults could trigger automatic changes in the control algorithms or operational scenario, depending on real-time operating limits and conditions.

Formation and sustainment of internal transport barriers in the International Thermonuclear Experimental Reactor with the baseline heating mix

Plasmas with internal transport barriers (ITBs) are a potential and attractive route to steady-state operation in ITER. These plasmas exhibit radially localized regions of improved confinement with steep pressure gradients in the plasma core, which drive large bootstrap current and generate hollow current profiles and negative magnetic shear. This work examines the formation and sustainment of ITBs in ITER with electron cyclotron heating and current drive. The time-dependent transport simulations indicate that, with a trade-off of the power delivered to the equatorial and to the upper launcher, the sustainment of steady-state ITBs can be demonstrated in ITER with the baseline heating configuration.

Physics of Plasma Control Toward Steady-State Operation of ITER

The ITER project is an important step toward fusion energy utilization for human beings. Fundamental understanding of physics is quite important, as well as understanding of tokamak systems and plasma control. In this lecture, given at the 4th ITER International Summer School, we give an introduction to tokamak research on fusion energy and ITER in light of the main theme of this school, plasma control, and we present some examples to illustrate the importance of physics by showing some physics elements underlying research toward steady-state operation of reactor-relevant tokamak plasmas.

AINA Safety Code, A Review of Loss of Plasma Control Transients in ITER: Sudden Increase in Fuelling Rate, Sudden Increase of Auxiliary Heating

In this contribution, two ITER loss of plasma control events are investigated. The first one is a failure of pumps and/or pellet injection system which fuel the plasma, producing an overfuelling event. The second examines an increase of external heating during the steady state operation of ITER. For both events, initial simulation parameters are scanned in order to find out which could lead to highest fusion power.Extensive simulation work has been done with AINA-1.0 (July-2007), the safety code developed by Fusion Energy Engineering Laboratory (FEEL-UPC) on the basis of SAFALY. AINA is a hybrid code comprising a zero-dimensional plasma dynamics and radial and poloidal thermal analyses of in-vessel components, and is intended to the quantitative investigation of plasma events in nuclear fusion reactors such as ITER.FEEL-UPC research group has been developing AINA code since 2004. AINA code brings significant improvements in relation to the original SAFALY code, and FEEL-UPC aims to continue improving it in the near future, with new models for plasma wall interactions, blanket thermal analysis, and plasma edge and core.

NBI: Progress in Physics and Engineering toward Fusion Reactors: N-NBI Heating and Current Drive in JT-60U towards Steady-State Tokamak

In steady-state tokamak fusion reactors, an efficient external current drive and a large fraction of the bootstrap current are required for non-inductive operation at low circulating power. NBI is a powerful and reliable actuator for the current drive and heating. A negative-ion based NBI (N-NBI) with a high beam energy more than 350 keV has been installed in the JT-60U tokamak in order to study the NBI current drive and heating in an ITER relevant regime. This paper presents recent achievements in N-NBI experiments and progress of the JT-60U N-NBI system for steady-state operation in ITER and tokamak fusion reactors.

Physics of transport barriers

The concept of ExB flow velocity shear suppression is utterly fundamental in modern fusion research. It is asserted that there are models enabling to understand the physics involved in LH transitions. To improve the understanding of the mechanisms leading to the formation of Transport Barriers, especially the relation between Internal and Edge barriers it is necessary to invoke the issue of electric fields. Edge transport barriers are the feature of the H-mode, the baseline regime of ITER, whereas Internal Transport Barriers are used to develop regimes that might be employed for steady state operation of ITER , definitely beneficial for design and operation of fusion power plants in the future. Their synergy will be addressed. Plasma are closely connected to electric fields. Therefore, their role is crucial for understanding of tokamaks aimed at the achievement of fusion energy. This appears in the well known neoclassical theory as the most accomplished and selfconsistent basis for understanding of fusion plasmas. It pertains to the novel concept of zonal flows emerging from the recent development of gyro-kinetic transport codes. The equilibrium poloidal and toroidal are also crucial for the concept of the electric field shear suppression of plasma turbulence in tokamaks. Yet, this timely and topical issue has remained largely unaddressed experimentally because of great difficulties in measuring in plasmas.

Overview of physics basis for ITER

ITER will be the first magnetic confinement device with burning DT plasma and fusion power of about 0.5 GW. Parameters of ITER plasma have been predicted using methodologies summarized in the ITER Physics Basis (1999 Nucl. Fusion 39 2175). During the past few years, new results have been obtained that substantiate confidence in achieving Q ⩾ 10 in ITER with inductive H-mode operation. These include achievement of a good H-mode confinement near the Greenwald density at high triangularity of the plasma cross section; improvements in theory-based confinement projections for the core plasma, even though further studies are needed for understanding the transport near the plasma edge; improvement in helium ash removal due to the elastic collisions of He atoms with D/T ions in the divertor predicted by modelling; demonstration of feedback control of neoclassical tearing modes and resultant improvement in the achievable β-values; better understanding of edge localized mode (ELM) physics and development of ELM mitigation techniques; and demonstration of mitigation of plasma disruptions. ITER will have a flexibility to operate also in steady-state and intermediate (hybrid) regimes. The ‘advanced tokamak’ regimes with weak or negative central magnetic shear and internal transport barriers are considered as potential scenarios for steady-state operation. The paper concentrates on inductively driven plasma performance and discusses requirements for steady-state operation in ITER.

Development of high-performance discharges with transport barriers in JT-60U

In order to provide greater confidence in the advanced steady-state operation in ITER, recent high-performance campaign in JT-60U has emphasized maximizing the capabilities of internal and edge transport barriers by the elaborate profile control using NB, electron cyclotron heating and other non-inductive current drive (CD) schemes as well as the strong shaping. Accordingly, it is highlighted by the reliable reproduction of the D–T equivalent fusion amplification gain (QDTeq) of over 1.2 transiently and sustainment of QDTeq at 0.8 for 0.55 s in reversed shear plasmas. On the other hand, increased triangularity up to 0.6 has also provided the improved confinement at higher density and sustainment of high normalized β (βN) above 2.7 for over 7 s in high-β poloidal H-mode. In addition, increased negative ion based neutral beam heating and CD power up to 5.7 MW contributed much to achieve the βN of 2.4 and confinement enhancement factor of 1.2 over the ITER98(y, 2) scaling simultaneously under full non-inductive CD at 1.8 MA. Furthermore, impact of electron heating on transport barriers was investigated, and it was found that strong barriers survive Te⩾Ti. With regard to the exploratory control of transport barriers using the perturbative schemes, the effect of asymmetric pressure fluctuations and Maxwell stress has been investigated. The suggested approaches could be applied in steady-state burning plasmas with transport barriers, where the external heating fraction is reduced and electron heating is predominant.

Pumping performance of pebble drop divertor

The pebble drop divertor concept was proposed for compact and high power density fusion reactors. In this concept, fuel gas can be pumped out by using fuel gas gettering by the surface layer of the pebbles. We are interested in the pumping by using transient hydrogen release from graphite just after ion irradiation. In this study, the pumping performance of the pebble drop divertor by using the transient hydrogen release is experimentally studied. In the experiment, the graphite samples were irradiated by high flux hydrogen beam with a maximum flux of 1×10 22 H/m 2 s (H 3 +, 5 keV). Just after the irradiation, hydrogen release from the samples was measured with keeping the sample temperature at the temperature reached by the irradiation. The transient hydrogen release was observed in all cases. The amount of the transient hydrogen release increased with increasing temperature. The amount of maximum transient hydrogen release was 4.4×10 20 H/m 2 at the temperature of 973 K. From this result, the pumping performance of the pebble drop divertor will satisfy the specification of fuel gas feed needed in steady state operation of ITER.