Bootstrap Fraction Research Articles

Variable-spectrum mode control of high poloidal beta discharges

DIII-D experiments demonstrate that high pressure, broad current profile equilibria can be accessed in the high poloidal beta regime by optimizing the MHD mode control poloidal spectrum. A novel, variable spectrum (VS) magnetic feedback scheme implemented using the DIII-D internal non-axisymmetric coils (I-coils) facilitated access to reduced internal inductance ℓi operation above the no-wall beta limit compared with both no feedback and fixed spectrum feedback. In addition, the VS feedback helped avoid beta collapses caused by marginally unstable resistive wall mode activity. The lower and upper I-coil rows were configured in two independent feedback loops, allowing the feedback field’s poloidal spectrum to vary and track changes in the plasma mode structure as the edge safety factor q 95 varied from 11 to 6 during the discharges. The q 95 dependence of the measured phase difference between the lower and upper I-coil rows during VS feedback is qualitatively compatible with ideal MHD simulations of the least-stable plasma kink mode and with plasma response simulations that included kinetic modifications to ideal MHD. The VS feedback approach is a straightforward way to improve resilience to variations in mode structure that occur as plasma parameters change. The demonstrated expansion of the operating space to lower ℓi is expected to improve the coupling of the plasma kink mode to external fields and beneficial wall eddy currents, and is compatible with high bootstrap fraction operation.

Efficient ECCD non-inductive plasma current start-up, ramp-up, and sustainment for an ST fusion reactor

The elimination of the need for an Ohmic heating solenoid may be the most impactful design driver for the realization of economical compact fusion tokamak reactor systems. However, this would require fully non-inductive start-up and current ramp-up from zero plasma current and low electron temperature of sub-keV to the full plasma current of ∼10–15 MA at 20–30 keV electron temperature. To address this challenge, an efficient solenoid-free start-up and ramp-up scenario utilizing a low-field-side-launched extraordinary mode at the fundamental electron cyclotron harmonic frequency (X–I) is proposed, which has more than two orders of magnitude higher electron cyclotron current drive (ECCD) efficiency than the conventional ECCD for the sub-keV start-up regime. A time dependent model was developed to simulate the start-up scenarios. For the Spherical Tokamak Advanced Reactor (STAR) (Menard et al 2023 Next-Step Low-Aspect-Ratio Tokamak Design Studies (IAEA)), it was found that to fully non-inductively ramp-up to 15 MA, it would take about 25 MW of EC power at 170 GHz. Because of the relatively large plasma volume of STAR, radiation losses must be considered. It is important to make sure that high Z impurities are kept sufficiently low during the early current start-up phase where the temperature is sub-keV range. Since the initial current ramp up takes place at a factor of ten lower density compared to the sustained regimes, it is important to transition into a higher bootstrap fraction discharge at lower density to minimize the ECCD power requirement during the densification. For the sustainment phase an array of eight gyrotron launchers with a total of about 60 MW of fundamental O-mode was found to be sufficient to provide the required axis-peaked external current drive. High efficiencies between 19–57 kA MW−1 were found with optimal aiming, and these were resilient to small changes in aiming angles and density and temperature profiles.

Parameter Space Constraints for Compact Spherical Tokamak Fusion Reactors

Parameter space for spherical tokamak reactors is explored quantitatively to elucidate the main constraints for spherical tokamak design choices. Using a constant plasma current Ip search constraint, a set of four Ip scenarios (5, 10, 15, and 20 MA) is first explored in a wide parameter space. Considering modest but gradually increasing auxiliary power, a set of four machine configurations (major radius R = 1.25, 1.75, 2.25, and 3.5 m) is explored next, optimizing the Ip and the bootstrap fraction. Constraints that narrow down the vast parameter space are elaborated along with critical assumptions, such as current drive efficiency, H-mode enhancement factor, nuclear shielding efficiency, and confinement scaling. Limits on the current density of the center post and how it affects the shielding are quantitatively indicated, thereby setting a lower limit on the aspect ratio.

An Integrated Design Study for an Advanced Tokamak to Close Physics Gaps in Energy Confinement and Power Exhaust

A high-level design study for a new experimental tokamak shows that advances in fusion science and engineering can be leveraged to narrow the gaps in energy confinement and exhaust power handling that remain between present devices and a future fusion pilot plant (FPP). This potential new U.S. facility, an Exhaust and Confinement Integration Tokamak Experiment (EXCITE), will access an operational space close to the projected FPP performance regime via a compact, high-field, high-power-density approach that utilizes advanced tokamak scenarios and high-temperature superconductor magnets. Full-device optimization via system code calculations, physics-based core-edge modeling, plasma control simulations, and finite element structural and thermal analysis has converged on a T, MA, m, , D-D tokamak with strong plasma shaping, long-legged divertors, and 50 MW of auxiliary power. Such a device will match several absolute FPP parameters: plasma pressure, exhaust heat flux, and toroidal magnetic field. It will also narrow or close the gap in key dimensionless parameters: toroidal beta, bootstrap fraction, collisionality, and edge neutral opacity. Integrated neutron shielding preserves personnel access by limiting nuclear activation and maximizes experimental run time by reducing site radiation. In addition to design study results and optimization details, parameter sensitivities and uncertainties are also discussed.

Characteristics of current density profile in EAST high poloidal beta operation

Current density profile optimization is one of the key issues to achieve high performance plasma operation. Much endeavor of current profile broadening has been made in EAST in past few years. This article thoroughly illustrates the characteristics of current density profile in EAST experiments. By summarizing the recent progress of the extension of the poloidal beta, it is found that the current density profile is broadened due to higher bootstrap fraction obtained with higher poloidal beta, which is often achieved under higher plasma density. Higher plasma density also helps shift current density and power deposition of lower hybrid wave outward from the plasma core according to analysis results, which accounts for the current profile broadening as well. Higher bootstrap fraction can be obtained under injection of neutral beam, thus further broadening current density profile. Limitation pertains to the fast ion pressure increasement under neutral beam in this strategy, requiring further optimization of fast ion pressure in the future.

Predictive modeling of operation scenarios with negative central magnetic shear and large-radius ITBs on EAST

Negative central magnetic shear or magnetic shear reversal is an attractive aspect for advanced operation scenarios in tokamaks as it suppresses turbulence and facilitates the formation of internal transport barriers (ITBs). The advanced operation scenario with negative magnetic shear, which is associated with high confinement quality, a large bootstrap fraction, ITBs, and so on, is one of the future goals of the experiments on EAST. In this work, modeling efforts have been made to find the operation regimes on EAST with negative central shear and large-radius ITBs. The modeling results indicate that negative central shear and large-radius ITBs can be achieved when the electron density is 〈ne〉/nG > 0.7 and the off-axis deposited electron cyclotron heating (ECH) power is 2 MW. Additionally, scenarios with lower ECH power (∼1.5 MW) have been studied since at present, the maximum ECH power that can be provided on EAST is 1.5 MW. With a lower off-axis ECH power of 1.5 MW, scenarios with negative central shear and large-radius ITBs but with a higher electron density of 〈ne〉/nG > 0.8 are obtained. The same as the first case, if the electron density is reduced to 〈ne〉/nG = 0.8 in this case, large-radius ITBs disappeared in the predicted scenario. This reveals that high electron density and large off-axis deposited ECH power are important for scenarios that have q-profiles with negative central shear and large-radius ITBs since lower hybrid current drive (LHCD) tends to peak near the axis at lower electron density, and hence, it helps to increase the temperature gradient near the axis, which will promote bootstrap in this region, leading to a centrally peaked total current density profile. Higher off-axis deposited ECH power helps to obtain scenarios with strong negative central shear and large-radius ITBs at a lower density. Scenarios predicted in this work will guide future experiments on EAST.

NSTX-U theory, modeling and analysis results

The mission of the low aspect ratio spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step steady-state tokamak fusion devices. NSTX-U will ultimately operate at up to 2 MA of plasma current and 1 T toroidal field on axis for 5 s, and has available up to 15 MW of neutral beam injection power at different tangency radii and 6 MW of high harmonic fast wave heating. With these capabilities NSTX-U will develop the physics understanding and control tools to ramp-up and sustain high performance fully non-inductive plasmas with large bootstrap fraction and enhanced confinement enabled via the low aspect ratio, high beta configuration. With its unique capabilities, NSTX-U research also supports ITER and other critical fusion development needs. Super-Alfvénic ions in beam-heated NSTX-U plasmas access energetic particle (EP) parameter space that is relevant for both α-heated conventional and low aspect ratio burning plasmas. NSTX-U can also generate very large target heat fluxes to test conventional and innovative plasma exhaust and plasma facing component solutions. This paper summarizes recent analysis, theory and modelling progress to advance the tokamak physics basis in the areas of macrostability and 3D fields, EP stability and fast ion transport, thermal transport and pedestal structure, boundary and plasma material interaction, RF heating, scenario optimization and real-time control.

NSTX-U theory, modeling and analysis results

Abstract The mission of the low aspect ratio spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step steady-state tokamak fusion devices. NSTX-U will ultimately operate at up to 2 MA of plasma current and 1 T toroidal field on axis for 5 seconds, and has available up to 15 MW of Neutral Beam Injection (NBI) power at different tangency radii and 6 MW of High Harmonic Fast Wave (HHFW) heating. With these capabilities NSTX-U will develop the physics understanding and control tools to ramp-up and sustain high performance fully non-inductive plasmas with large bootstrap fraction and enhanced confinement enabled via the low aspect ratio, high beta configuration. With its unique capabilities, NSTX-U research also supports ITER and other critical fusion development needs. Super-Alfvénic ions in beam-heated NSTX-U plasmas access energetic particle parameter space that is relevant for both -heated conventional and low aspect ratio burning plasmas. NSTX-U can also generate very large target heat fluxes to test conventional and innovative plasma exhaust and plasma facing component (PFC) solutions. This paper summarizes recent analysis, theory and modelling progress to advance the tokamak physics basis in the areas of macrostability and 3D fields, energetic particle stability and fast ion transport, thermal transport and pedestal structure, boundary and plasma material interaction, RF heating, scenario optimization and real-time control.

Enhanced pedestal H-mode at low edge ion collisionality on NSTX

The Enhanced Pedestal (EP) H-mode regime is an attractive wide-pedestal high-βp scenario for the National Spherical Torus Experiment Upgrade (NSTX-U) and next-step devices as it achieves enhanced energy confinement (H98y,2 > 1.5), large normalized pressure (βN > 5), and significant bootstrap fraction (fBS > 0.6) at Ip/BT = 2 MA/T. This regime is realized when the edge ion collisionality becomes sufficiently small that a positive feedback interaction occurs between a reduction in the ion neoclassical energy transport and an increase in the particle transport from pressure-driven edge instabilities. The EP H-mode was most often observed as a transition following a large edge-localized mode in conditions with low edge neutral recycling. It is hypothesized that the onset of pressure-driven instabilities prior to the full recovery of the neutral density leads to a temporary period with an elevated ion temperature gradient that triggers the transition to EP H-mode. Linear CGYRO and M3D-C1 calculations are compared to beam emission spectroscopy and magnetic spectroscopy in order to describe the evolution of the edge particle transport mechanisms during the ELM recovery and the saturated EP H-mode state. The observations are consistent with the hypothesis that the onset of pressure-driven edge instabilities, such as the kinetic ballooning mode and kink-peeling mode, can be responsible for the increased particle transport in EP H-mode.

Improved high-performance fully non-inductive discharge by optimizing the fast-ion confinement on EAST

The attainment of long-pulse, high-performance, fully non-inductive plasma is one of the major scientific objectives of EAST, using the ITER-like tungsten upper divertor. Understanding and optimizing the fast-ion behaviors is the critical issue to extending the performance of EAST. Recently, using both neutral beam injection (NBI) and radio frequency (RF; low hybrid, electron cyclotron, and ion cyclotron) heating, fully non-inductive high-βP scenarios with extension of fusion performance at high density and low rotation have been achieved, with βP up to 2.5, βN up to 2.0, H98y 2 > 1.1, and bootstrap current fraction (f BS) up to 50%. For previous long-pulse H-mode plasma at medium density, when NBI is added into RF plasma, βp is increased from 1.2 to 2.0 compared with RF-only discharges. In fact, f BS for both discharges is nearly the same, at ~22%. Analysis shows that the increase in βp is mostly due to fast ions which do not contribute significantly to the neoclassical bootstrap current. Thus, to obtain high-performance plasmas with improved bootstrap current fraction, key parameters (e.g. density, beam energy, etc.) must be further optimized. Experimental results show that high density improves bootstrap fraction also by reducing fast-ion slowing-down time and loss. The lower beam energy also mitigates fast-ion loss, which is better for heating and CD performance. The extension of high-performance, fully non-inductive experiments on EAST at high density and zero/low NBI torque can potentially offer unique contributions towards ITER and CFETR.

Modeling and advances in the high bootstrap fraction regime on EAST towards the steady-state operation

Experimental and modeling investigations on the Experimental Advanced Superconducting Tokamak (EAST) show attractive confinement and stability properties in fully non-inductive, high poloidal beta plasmas. In the 2018 EAST experimental campaign, extended operation regimes of steady-state scenario were obtained (βP ~ 1.9 & βN ~ 1.5 & H98y 2 ~ 1.3 of using only RF heating) with a high bootstrap current fraction (f BS ~ 47%) and ne/nGW ~ 70%. The confinement quality, H98y 2 ~ 1.3, is much better than standard H-mode, and stationary peaked electron temperature profiles and peaked current density profile when ~1 MW of ECH and ~2.6 MW of LHW are both deposited in the core region. The observed improvement in plasma confinement is much better (H98y 2 ~ 1.3) when compared with the RF-dominant heating experiments in the EAST 2016–2017 experimental campaign (H98y 2 ~ 1.1). Integrated modeling prediction suggests that high electron density would increase the plasma performance and bootstrap current fraction, which is consistent with the general experimental trend. Linear analysis shows that the high-k (ky > 1) modes instability (ETG) is suppressed in the core region. Also, the Shafranov shift is shown to play a role in the suppression of the electron turbulent energy transport. Besides the modeling predictions, the validation of the predicted of the effect of ECH on the plasma confinement in recent experiments was done and the experimental results were consistent with the modeling results. The validation results also suggest that when ECH is deposited in the core region in the RF heating experiments, increasing the ECH heating power from 0.5 MW to 1.0 MW does make a small improvement in the bootstrap current fraction. The high bootstrap fraction scenario realized on EAST and the investigation to achieve higher-performance plasma would help expanding the operation regime on EAST.

Recent advances in EAST physics experiments in support of steady-state operation for ITER and CFETR

Since the last IAEA Fusion Energy Conference in 2016, the EAST physics experiments have been developed further in support of high-performance steady-state operation for ITER and CFETR. First demonstration of a >100 s time scale long-pulse steady-state scenario with a good plasma performance (H98(y2) ~ 1.1) and a good control of impurity and heat exhaust with the upper tungsten divertor has been achieved on EAST using the pure radio frequency (RF) power heating and current drive. The EAST operational domain has been significantly extended towards a more ITER and CFETR related high beta steady-state regime (βP ~ 2.5 and βN ~ 1.9 of using RF and NB and βP ~ 1.9 and βN ~ 1.5 of using pure RF). A large bootstrap current fraction up to 47% has been achieved with with q95 ~ 6.0–7.0. The interaction effect between the electron cyclotron resonant heating and two lower hybrid wave systems has been investigated systematically, and applied for the improvement of current drive efficiency and plasma confinement quality in the steady-state scenario development on EAST. Full edge-localized mode (ELM) suppression using the n = 2 resonant magnetic perturbations has been achieved in ITER-like standard type-I ELMy H-mode plasmas with a range of the edge safety factor of q95 ≈ 3.2–3.7 on EAST. Reduction of the peak heat flux on the divertor was demonstrated using the active radiation feedback control. An increase in the total heating power and improvement of the plasma confinement are expected using a 0D model prediction for a higher bootstrap fraction. Towards a long-pulse, high bootstrap current fraction operation, a new lower ITER-like tungsten divertor with active water-cooling will be installed, together with further increase and improvement of heating and current drive capability.

Theory of transport in high bootstrap fraction H-modes with internal transport barriers

A review of the recent progress in understanding the transport in fully non-inductive, high bootstrap fraction H-mode discharges, developed on the DIII-D tokamak, is given in this paper. Experiments have demonstrated improved energy confinement due to the formation of an internal transport barrier in high bootstrap fraction H-mode discharges. Gyrokinetic analysis, and quasi-linear predictive modeling, demonstrates that the observed transport barrier is due to the suppression of turbulence, primarily due to the large Shafranov shift. E × B velocity shear does not play a significant role in the transport barrier due to the high safety factor. The ion energy transport is reduced to neoclassical for DIII-D discharges and electron energy and particle transport is reduced, but still turbulent, within the barrier. Deeper into the plasma, high levels of electron transport are observed. The electron energy transport in DIII-D is found to be due to electromagnetically enhanced electron scale modes. A contribution from fast ion driven instabilities is also possible. The kinetic ballooning mode plays a strong role in the region between the internal and H-mode edge transport barriers with two types of self-organized pedestal states observed.

Transport barriers in bootstrap-driven tokamaks

Experiments have demonstrated improved energy confinement due to the spontaneous formation of an internal transport barrier in high bootstrap fraction discharges. Gyrokinetic analysis, and quasilinear predictive modeling, demonstrates that the observed transport barrier is caused by the suppression of turbulence primarily from the large Shafranov shift. It is shown that the Shafranov shift can produce a bifurcation to improved confinement in regions of positive magnetic shear or a continuous reduction in transport for weak or negative magnetic shear. Operation at high safety factor lowers the pressure gradient threshold for the Shafranov shift-driven barrier formation. Two self-organized states of the internal and edge transport barrier are observed. It is shown that these two states are controlled by the interaction of the bootstrap current with magnetic shear, and the kinetic ballooning mode instability boundary. Election scale energy transport is predicted to be dominant in the inner 60% of the profile. Evidence is presented that energetic particle-driven instabilities could be playing a role in the thermal energy transport in this region.

Transport modeling of the DIII-D high scenario and extrapolations to ITER steady-state operation

Transport modeling of a proposed ITER steady-state scenario based on DIII-D high poloidal-beta () discharges finds that ITB formation can occur with either sufficient rotation or a negative central shear q-profile. The high scenario is characterized by a large bootstrap current fraction (80%) which reduces the demands on the external current drive, and a large radius internal transport barrier which is associated with excellent normalized confinement. Modeling predictions of the electron transport in the high scenario improve as approaches levels similar to typical existing models of ITER steady-state and the ion transport is turbulence dominated. Typical temperature and density profiles from the non-inductive high scenario on DIII-D are scaled according to 0D modeling predictions of the requirements for achieving a steady-state fusion gain in ITER with ‘day one’ heating and current drive capabilities. Then, TGLF turbulence modeling is carried out under systematic variations of the toroidal rotation and the core q-profile. A high bootstrap fraction, high scenario is found to be near an ITB formation threshold, and either strong negative central magnetic shear or rotation in a high bootstrap fraction are found to successfully provide the turbulence suppression required to achieve .

Advances in the high bootstrap fraction regime on DIII-D towards the Q = 5 mission of ITER steady state

Recent EAST/DIII-D joint experiments on the high poloidal beta regime in DIII-D have extended operation with internal transport barriers (ITBs) and excellent energy confinement (H98y2 ~ 1.6) to higher plasma current, for lower q95 ⩽ 7.0, and more balanced neutral beam injection (NBI) (torque injection < 2 Nm), for lower plasma rotation than previous results (Garofalo et al, IAEA 2014, Gong et al 2014 IAEA Int. Conf. on Fusion Energy). Transport analysis and experimental measurements at low toroidal rotation suggest that the E × B shear effect is not key to the ITB formation in these high discharges. Experiments and TGLF modeling show that the Shafranov shift has a key stabilizing effect on turbulence. Extrapolation of the DIII-D results using a 0D model shows that with the improved confinement, the high bootstrap fraction regime could achieve fusion gain Q = 5 in ITER at ~ 2.9 and q95 ~ 7. With the optimization of q(0), the required improved confinement is achievable when using 1.5D TGLF-SAT1 for transport simulations. Results reported in this paper suggest that the DIII-D high scenario could be a candidate for ITER steady state operation.

Compatibility of internal transport barrier with steady-state operation in the high bootstrap fraction regime on DIII-D

Recent EAST/DIII-D joint experiments on the high poloidal beta tokamak regime in DIII-D have demonstrated fully noninductive operation with an internal transport barrier (ITB) at large minor radius, at normalized fusion performance increased by ⩾30% relative to earlier work (Politzer et al 2005 Nucl. Fusion 45 417). The advancement was enabled by improved understanding of the ‘relaxation oscillations’, previously attributed to repetitive ITB collapses, and of the fast ion behavior in this regime. It was found that the ‘relaxation oscillations’ are coupled core-edge modes amenable to wall-stabilization, and that fast ion losses which previously dictated a large plasma-wall separation to avoid wall over-heating, can be reduced to classical levels with sufficient plasma density. By using optimized waveforms of the plasma-wall separation and plasma density, fully noninductive plasmas have been sustained for long durations with excellent energy confinement quality, bootstrap fraction ⩾80%, , , and . These results bolster the applicability of the high poloidal beta tokamak regime toward the realization of a steady-state fusion reactor.

ARC: A compact, high-field, fusion nuclear science facility and demonstration power plant with demountable magnets

The affordable, robust, compact (ARC) reactor is the product of a conceptual design study aimed at reducing the size, cost, and complexity of a combined fusion nuclear science facility (FNSF) and demonstration fusion Pilot power plant. ARC is a ∼200–250MWe tokamak reactor with a major radius of 3.3m, a minor radius of 1.1m, and an on-axis magnetic field of 9.2T. ARC has rare earth barium copper oxide (REBCO) superconducting toroidal field coils, which have joints to enable disassembly. This allows the vacuum vessel to be replaced quickly, mitigating first wall survivability concerns, and permits a single device to test many vacuum vessel designs and divertor materials. The design point has a plasma fusion gain of Qp≈13.6, yet is fully non-inductive, with a modest bootstrap fraction of only ∼63%. Thus ARC offers a high power gain with relatively large external control of the current profile. This highly attractive combination is enabled by the ∼23T peak field on coil achievable with newly available REBCO superconductor technology. External current drive is provided by two innovative inboard RF launchers using 25MW of lower hybrid and 13.6MW of ion cyclotron fast wave power. The resulting efficient current drive provides a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing fluorine lithium beryllium (FLiBe) molten salt. The liquid blanket is low-risk technology and provides effective neutron moderation and shielding, excellent heat removal, and a tritium breeding ratio≥1.1. The large temperature range over which FLiBe is liquid permits an output blanket temperature of 900K, single phase fluid cooling, and a high efficiency helium Brayton cycle, which allows for net electricity generation when operating ARC as a Pilot power plant.

Designing a tokamak fusion reactor—How does plasma physics fit in?

This paper attempts to bridge the gap between tokamak reactor design and plasma physics. The analysis demonstrates that the overall design of a tokamak fusion reactor is determined almost entirely by the constraints imposed by nuclear physics and fusion engineering. Virtually, no plasma physics is required to determine the main design parameters of a reactor: a, R0, B0, Ti, Te, p, n, τE, I. The one exception is the value of the toroidal current I, which depends upon a combination of engineering and plasma physics. This exception, however, ultimately has a major impact on the feasibility of an attractive tokamak reactor. The analysis shows that the engineering/nuclear physics design makes demands on the plasma physics that must be satisfied in order to generate power. These demands are substituted into the well-known operational constraints arising in tokamak physics: the Troyon limit, Greenwald limit, kink stability limit, and bootstrap fraction limit. Unfortunately, a tokamak reactor designed on the basis of standard engineering and nuclear physics constraints does not scale to a reactor. Too much current is required to achieve the necessary confinement time for ignition. The combination of achievable bootstrap current plus current drive is not sufficient to generate the current demanded by the engineering design. Several possible solutions are discussed in detail involving advances in plasma physics or engineering. The main contribution of the present work is to demonstrate that the basic reactor design and its plasma physics consequences can be determined simply and analytically. The analysis thus provides a crisp, compact, logical framework that will hopefully lead to improved physical intuition for connecting plasma physic to tokamak reactor design.

Understanding of the fundamental differences in JET and JT-60U AT discharges

Plasma current density simulations of JET and JT-60U shots in reverse-q advanced scenarios based on the previous data analysis of the identity plasma experiments have been performed. The effects of the main differences between the shots (neutral beam current density (jnbi), electron density and geometry) have been studied. The reversed q-profile (which was the target in this identity experiment and is observed at the beginning of each shot) was sustained in JT-60U while it became flat in JET towards the end of the shot. In JET, jnbi is peaked on-axis whereas in JT-60U it is peaked off-axis (at ρ = 0.5) while the NBI fraction of the selected shots is the same (22–24%). A strong density ITB appeared (at ρ = 0.5) in JT-60U but not in JET. The plasma geometry was mainly set to match but it was not identical. In addition, the extrapolation to JET steady-state operation has been done by testing the sensitivity of q to different externally driven currents, electron density and geometry in predictive current diffusion simulations. Moreover, critical bootstrap current density has been analysed.The reasons for the different time evolution of q-profile have been studied with predictive current simulations with the 1.5D transport code JETTO. The current diffusion model was validated against reverse-q shots, and simulations were performed with experimental data profiles, jNBI given by ASCOT and neoclassical resistivity and bootstrap current calculated by NCLASS.Bootstrap current density was the most efficient way to sustain the beneficial reverse shape of q-profile. Replacing the JET ne profile with one from JT-60U leads to an increase of 0.2–0.3 MA in the bootstrap current (f bs increases from 15% to 30%). However, sustaining the stationary reverse q is not achieved in JET with bootstrap current induced by the density gradient of JT-60U. Even 10 times larger gradient than in JT-60U helps to sustain the shape of the q-profile longer than the experimentally observed density profile in JET but the minimum value of q moves closer to central plasma and the shape of q is not stationary in a 10 s simulation. The effect of different shape of NBI current density profile is negligible due to quite small fraction. Sustaining the reverse q requires 45% or larger added off-axis fraction. However, increasing the inverse aspect ratio increases bootstrap current and decreases the critical bootstrap current more effectively than increasing the density gradient: two times larger inverse aspect ratio produces almost three times larger bootstrap fraction but ten times larger density gradient only two times larger bootstrap fraction. The conclusions based on these simulations indicate that the need for bootstrap current is larger in JET and the same conditions cause smaller bootstrap fraction, which suggests that achieving steady-state operation in JET under these conditions is unlikely.