Thermal Energy Confinement Time Research Articles

Development of high-performance long-pulse discharge in KSTAR

High-performance long-pulse plasma operation is essential for producing economically viable fusion energy in tokamak devices. To achieve such discharges in KSTAR, firstly, the rapid increase in the temperature of plasma-facing components was mitigated. The temperature increase of the poloidal limiter, especially, was associated with beam-driven fast ion orbit loss and the discrepancy of the equilibrium reconstructed with heated magnetic probes of signal drift. The fast ions lost to the poloidal limiter were reduced by optimizing the plasma shape and the composition of neutral beam injection (NBI). This nonlinear signal drift was successfully reduced by a new thermal shielding protector on the magnetic probes. Secondly, a lower loop voltage approach was implemented to reduce a poloidal flux consumption rate. A plasma current of 400 kA and a line-averaged electron density of ∼2.0 × 1019 m−3 were chosen by considering the L–H power threshold, fast ion orbit loss, and beam shine-through power loss for low loop voltage in KSTAR. In addition, the application of electron cyclotron heating also helped maintain the plasma with low loop voltage (∼25 mV) by enhancing the NBI-driven current and achieving a high poloidal beta (β P) state. KSTAR has achieved a long pulse (∼90 s) operation with the high performance of β P ⩽ 2.7, thermal energy confinement enhancement factor (H98y2) ∼ 1.1, and fraction of non-inductive current (f NI) ⩽ 0.96. Still, gradual degradation of the plasma performance has been observed over time in the discharges. In one of the long-pulse discharges, β P reduced by ∼18% over the time of ∼8τ R (current relaxation time, τ R ∼ 5 s) and ∼1067τ E,th (thermal energy confinement time, τ E,th ∼ 45 ms). The degradation may be closely associated with weak, yet growing, and persistent toroidal Alfvén eigenmodes and their effect on fast ion confinement.

The Formation and Stationary Maintenance of Peaked Pressure Profiles in Turbulent Tokamak Plasmas

The theoretical model for increasing peaking of radial distribution of plasma thermal energy and plasma energy confinement time improvement due to the formation of current profiles those provide the increased value of the safety factor qb at the outer plasma boundary is considered. The model for the formation and stationary maintenance of the peaked current density profiles those ensure the maintenance of plasma column equilibrium state with increased qb value and conservation of q0=1 at the magnetic axis is developed. The model is based on the significant ramp down of the Iind current maintained by the inductor with simultaneous central ECR heating and electron cyclotron current drive (ECCD) switching on. It is shown by the computer simulations of the self-consistent time evolution of turbulent plasma for the conditions of T-10 and T-15MD tokamaks that in the shot scenarios with three- or four-fold decrease in the Iind current at the ECRH stage of shots it was achieved an approximately two-fold enhancement of the electron and ion temperatures at the axis of the plasma column as compared to the standard shots with the same heating powers.

The Formation and Stationary Maintenance of Peaked Pressure Profiles in Turbulent Tokamak Plasmas

The theoretical model for increasing peaking of radial distribution of plasma thermal energy and plasma energy confinement time improvement due to the formation of current profiles those provide the increased value of the safety factor {{q}_{b}} at the outer plasma boundary is considered. The model for the formation and stationary maintenance of the peaked current density profiles those ensure the maintenance of plasma column equilibrium state with increased {{q}_{b}} value and conservation of {{q}_{0}} approx 1 at the magnetic axis is developed. The model is based on the significant ramp down of the {{I}_{{{text{ind}}}}} current maintained by the inductor with simultaneous central ECR heating and electron cyclotron current drive (ECCD) switching on. It is shown by the computer simulations of the self-consistent time evolution of turbulent plasma for the conditions of T-10 and T-15MD tokamaks that in the shot scenarios with three- or four-fold decrease in the {{I}_{{{text{ind}}}}} current at the ECRH stage of shots it was achieved an approximately two-fold enhancement of the electron and ion temperatures at the axis of the plasma column as compared to the standard shots with the same heating powers.

The scalings of the thermal energy confinement time in EAST H-mode plasmas

This paper describes scalings of the H-mode confinement database of neutral beam injection (NBI)-lower hybrid wave (LHW) and electron cyclotron resonant heating-LHW plasmas in the EAST tokamak. Fundamental information of the EAST H-mode database and the details of calculation of the thermal energy confinement time in EAST are presented. The result of the scaling model of the thermal energy confinement time in NBI-LHW plasmas with root-mean-square-error , compared with the multi-machine scaling model derived from the multi-machine database (mainly NBI heating) with , is obtained. The scaling presents a similar dependency on plasma current and power loss and a lower dependency on plasma density and elongation, but stronger dependency on toroidal field than in . The scaling demonstrates a similar dependency on plasma current and power loss and lower dependency on plasma density, but a negative dependency with large error bars on elongation compared with in NBI-LHW plasmas. The comparison of EAST models with the more recent multi-tokamak scaling ITPA20-IL shows a similar engineering dependency on plasma current, power loss and plasma density, but opposite dependency on toroidal field.

Review of hydrogen isotope effects on H-mode confinement in JT-60U

Hydrogen isotope effects on H-mode confinement in JT-60U are reviewed. The thermal energy confinement time becomes longer in deuterium by a factor of ∼1.4 than in hydrogen at a given absorbed power. When the absorbed power is fixed, the values of electron and ion temperature become evidently higher in deuterium than in hydrogen throughout the entire range of minor radius whereas the electron density profiles are nearly the same. Hence, the effective heat diffusivity becomes relatively lower in deuterium than in hydrogen. When the thermal stored energy is fixed, identical profiles for the electron density, electron temperature, and ion temperature are obtained for hydrogen and deuterium plasmas, whereas the required power clearly increases for hydrogen, which results in reduction of heat diffusivity for deuterium. The inverse of the ion-temperature-gradient (ITG) scale length required for a given ion heat diffusivity increased by a factor of approximately 1.2 for deuterium compared with that for hydrogen. Similar density profiles are formed for the hydrogen and deuterium plasmas despite the higher turbulent particle flux for the hydrogen case, suggesting that turbulent particle transport is more suppressed in the deuterium plasmas. The gyrokinetic calculation indicates that the stabilization effect of e-i collisionality on the TEM is more significant for deuterium plasmas and causes the difference in the quasilinear diffusivity between the two cases. The edge-core interplay leading to the overall H-mode confinement is also addressed where the effect of the hydrogen isotopic composition was involved. The correlation between edge pedestal pressure and global poloidal beta holds true consistently regardless of the difference in the isotopic composition. A higher value of global poloidal beta was obtained for deuterium because of its smaller ITG scale length and because of the additional stored energy in the thermal and fast ion components, the latter due to an increase of slowing down time with an increase of isotopic mass.

Fusion performance of spherical and conventional tokamaks: implications for compact pilot plants and reactors

Spherical tokamaks (STs) have features that make them a potentially attractive option for fusion power production compared to conventional tokamaks (CTs) including operation at high beta and high self-driven ‘bootstrap’ current. The thermal energy confinement time (τ Ε) also typically has a stronger dependence on toroidal magnetic field and a weaker dependence on plasma current, but so far it has not been established how this difference impacts performance under reactor conditions. This aspect is explored in this paper. Using empirical data from NSTX and MAST, and from multiple CTs, we investigate analytically and by using established fusion codes the potential fusion performance, characterised by the fusion triple product, nTτ Ε, and fusion power gain, Q fus, where n and T are the density and temperature respectively. We find that for similar values of field and fusion power, but smaller volume, STs can have nTτ Ε up to a factor of three higher and Q fus an order of magnitude higher than CTs. We identify the origin of this enhanced performance and outline a measurement to advance this finding. Potentially our results open an alternative and faster route to fusion power based on relatively small, low power STs.

Energy and particle confinement in JET H-mode plasma

This work describes the behaviour of the global energy and particle confinement on JET observed in a massive database of H-mode plasmas covering almost the whole lifetime of JET operations, both with the carbon and metal wall. The analysis is focused on type I ELMy H-modes in stationary phases. It is shown that plasma density in that regime is determined mainly by the plasma current, edge safety factor, triangularity of the last closed flux surface and the hydrogenic isotope mass. Behaviour is consistent for the whole database regardless of divertor configuration or the plasma facing materials. On average, thermal energy confinement time in JET with the carbon wall (JET-C) is accurately predicted by the IPB98(y,2) scaling. For JET with the ITER-like wall (JET-ILW), the energy confinement is found to be lower than expected from the scaling. The difference is found to be much stronger in a divertor configuration with the outer strike point at a vertical target and the pumping throat at the private flux. Observed lower confinement in JET-ILW can partially be attributed to the additional operational constraints of the metal wall machine, i.e. the avoidance of heavy impurity accumulation via additional gas dosing, but not fully. The isotope effect on the energy confinement in the Meff = 1–2 range is found to be the same in JET-C and JET-ILW, thus is independent of the wall material, if correlation with plasma density is accounted for. The effect of the toroidal magnetic field on the confinement is between zero and slightly negative. Triangularity has a generally favourable effect on the energy confinement, but the magnitude changes across the database. The effect of triangularity on plasma density is always much stronger than on energy confinement however, therefore plasmas with high triangularity are in general more dense and colder than at lower triangularities. The work described in this paper is completed under the EUROfusion global confinement database project, and the data shown here will be available to EUROfusion collaborators shortly. The data are also either already included in the international H-mode confinement database, or will be in the future updated version.

Evaluation of NBI absorption and fast ion stored energy to improve thermal energy confinement time calculation in EAST

The absorption of neutral beam power and the fast ion stored energy in EAST plasmas with neutral beam injection (NBI) is analyzed to improve the calculation of thermal energy confinement time. The neutral beam power absorption and fast ion stored energy are systematically calculated using the TRANSP code, through the investigation of global parameters including plasma current, line averaged density and beam energy. Results have shown that scaling laws for the NBI absorption coefficient and fast ion energy rate are obtained through statistical analysis. A comparison of the confinement improvement factor with these new scaling laws against those assuming fixed coefficients is given.

Isotope Effect on Energy Confinement Time and Thermal Transport in Neutral-Beam-Heated Stellarator-Heliotron Plasmas.

The isotope effect on energy confinement time and thermal transport has been investigated for plasmas confined by a stellarator-heliotron magnetic field. This is the first detailed assessment of an isotope effect in a stellarator heliotron. Hydrogen and deuterium plasmas heated by neutral beam injection on the Large Helical Device have exhibited no significant dependence on the isotope mass in thermal energy confinement time, which is not consistent with the simple gyro-Bohm model. A comparison of thermal diffusivity for dimensionally similar hydrogen and deuterium plasmas in terms of the gyroradius, collisionality, and thermal pressure has clearly shown robust confinement improvement in deuterium to compensate for the unfavorable mass dependence predicted by the gyro-Bohm model.

Isotope identity experiments in JET-ILW with H and D L-mode plasmas

NBI-heated L-mode plasmas have been obtained in JET with the Be/W ITER-like wall (JET-ILW) in H and D, with matched profiles of the dimensionless plasma parameters, ρ*, ν*, β and q in the plasma core confinement region and same Ti/Te and Zeff. The achieved isotope identity indicates that the confinement scale invariance principle is satisfied in the core confinement region of these plasmas, where the dominant instabilities are Ion Temperature Gradient (ITG) modes. The dimensionless thermal energy confinement time, Ωi τE,th, and the scaled core plasma heat diffusivity, A χeff/BT, are identical in H and D within error bars, indicating lack of isotope mass dependence of the dimensionless L-mode thermal energy confinement time in JET-ILW. Predictive flux driven simulations with JETTO-TGLF of the H and D identity pair is in very good agreement with experiment for both isotopes: the stiff core heat transport, typical of JET-ILW NBI heated L-modes, overcomes the local gyro-Bohm scaling of gradient-driven TGLF, explaining the lack of isotope mass dependence in the confinement region of these plasmas. The effect of E × B shearing on the predicted heat and particle transport channels is found to be negligible for these low beta and low momentum input plasmas.

On the energy confinement time in spherical tokamaks: implications for the design of pilot plants and fusion reactors

Experiments on NSTX and MAST have shown the thermal energy confinement time in spherical tokmaks (STs), , to have a stronger toroidal field and weaker plasma current dependence than in conventional large aspect ratio tokamaks. These scalings were derived for single machines both of which are similarly sized, consequently the NSTX and MAST scaling laws do not include a size dependence, and so cannot be used to extrapolate the performance of future STs. Using physics-based dimensional arguments we extend the NSTX scaling to include a size scaling. We also resolve a colinearity problem specific to NSTX data by assuming core transport is gyro-Bohm like. The resulting scaling has approximately zero beta dependence, a typical collisionality dependence, and a relatively weak safety factor dependence: . With the exception of the safety factor, all exponents are consistent with recent experiments in large aspect ratio tokamaks. This apparent difference between STs and large aspect ratio tokamaks is consistent with MAST and NSTX results. We have considered the implications of the scaling for pilot plants and reactors and find it may be possible to develop more compact reactors based on the ST approach.

Assessment of the baseline scenario at q95 ~ 3 for ITER

The International Tokamak Physics Activity topical group on integrated operational scenarios has compiled a database of stationary H-mode discharges at q95 ~ 3 from AUG, C-Mod, DIII-D, JET and JT-60U, for both carbon wall and high-Z metal wall experiments with ~3300 entries. The analyses focus on discharges that are stationary for ⩾5 thermal energy confinement times to evaluate the baseline scenario proposed for ITER at 15 MA for achieving its goals of Q = 10, fusion power of 500 MW at normalised pressure, βN = 1.8 and normalised confinement as predicted by the standard H-mode scaling, H98y2 = 1. With the data restricted to stationary H-modes at q95 ~ 3, the database shows significant variation of thermal energy confinement compared to the standard H-mode scaling (IPB98(y,2)) in dimensionless form. The data show similar scaling with normalised gyro-radius, but more favourable scaling towards lower collision frequency and more favourable scaling with plasma beta. Using all the engineering variables employed in IPB98(y,2), results in an overfit due to correlations among the data. Moreover, there are significant residual trends in the confinement for plasma current, device size, loss power, and in particular for the plasma density. Significant differences between results obtained for devices with a carbon wall and high-Z metal wall are observed in the data, with data from carbon wall devices providing a larger operating space, encompassing ITER parameters or even exceeding them. H-modes in high-Z metal wall devices have, so-far, not accessed conditions at low collision frequencies, have lower normalised confinement (H98y2 ~ 0.8–0.9) at low input power or beta, achieving H98y2 ~ 1.0 only at input powers two times the L- to H-mode transition scaling predictions and at βN ~ 2.0. Hence, only the best H-modes with high-Z metal walls reach ITER baseline performance requirements. The data show that operating at high plasma density, with line-averaged density at 85% of Greenwald density is achievable for H98y2 > 0.95 for a range of plasma configurations, and that operation at low plasma inductance with li(3) ~ 0.7–0.75 is feasible. Scenario simulations employed for projecting the plasma performance in ITER should incorporate a lower thermal confinement at low plasma beta for the entry to burn and provide projections using higher levels of plasma core radiation by plasma impurities. Moreover, ITER projections should not subtract the core radiation in the evaluation of the thermal confinement time and H98y2, to allow a fair comparison with experimental data currently available. From the data presented here, it is likely that in ITER the energy confinement time will not increase with plasma density and will have no degradation with plasma beta. The analyses indicate that the data at q95 ~ 3 are consistent with achievement of the ITER mission goals at 15 MA.

On the universality of power laws for tokamak plasma predictions

Significant deviations from well established power laws for the thermal energy confinement time, obtained from extensive databases analysis as the IPB98(y,2), have been recently reported in dedicated power scans. In order to illuminate the adequacy, validity and universality of power laws as tools for predicting plasma performance, a simplified analysis has been carried out in the framework of a minimal modeling for heat transport which is, however, able to account for the interplay between turbulence and collinear effects with the input power known to play a role in experiments with significant deviations from such power laws. Whereas at low powers, the usual scaling laws are recovered with little influence of other plasma parameters, resulting in a robust power low exponent, at high power it is shown how the exponents obtained are extremely sensitive to the heating deposition, the q-profile or even the sampling or the number of points considered due to highly non-linear behavior of the heat transport. In particular circumstances, even a minimum of the thermal energy confinement time with the input power can be obtained, which means that the approach of the energy confinement time as a power law might be intrinsically invalid. Therefore plasma predictions with a power law approximation with a constant exponent obtained from a regression of a broad range of powers and other plasma parameters which can non-linearly affect and suppress heat transport, can lead to misleading results suggesting that this approach should be taken cautiously and its results continuously compared with modeling which can properly capture the underline physics, as gyrokinetic simulations.

Global and pedestal confinement and pedestal structure in dimensionless collisionality scans of low-triangularity H-mode plasmas in JET-ILW

A dimensionless collisionality scan in low-triangularity plasmas in the Joint European Torus with the ITER-like wall (JET-ILW) has been performed. The increase of the normalized energy confinement (defined as the ratio between thermal energy confinement and Bohm confinement time) with decreasing collisionality is observed. Moreover, at low collisionality, a confinement factor H98, comparable to JET-C, is achieved. At high collisionality, the low normalized confinement is related to a degraded pedestal stability and a reduction in the density-profile peaking.The increase of normalized energy confinement is due to both an increase in the pedestal and in the core regions. The improvement in the pedestal is related to the increase of the stability. The improvement in the core is driven by (i) the core temperature increase via the temperature-profile stiffness and by (ii) the density-peaking increase driven by the low collisionality.Pedestal stability analysis performed with the ELITE (edge-localized instabilities in tokamak equilibria) code has a reasonable qualitative agreement with the experimental results. An improvement of the pedestal stability with decreasing collisionality is observed. The improvement is ascribed to the reduction of the pedestal width, the increase of the bootstrap current and the reduction of the relative shift between the positions of the pedestal density and pedestal temperature.The EPED1 model predictions for the pedestal pressure height are qualitatively well correlated with the experimental results. Quantitatively, EPED1 overestimates the experimental pressure by 15–35%. In terms of the pedestal width, a correct agreement (within 10–15%) between the EPED1 and the experimental width is found at low collisionality. The experimental pedestal width increases with collisionality. Nonetheless, an extrapolation to low-collisionality values suggests that the width predictions from the KBM constraint are reasonable for ITER.

Access to a new plasma edge state with high density and pressures using the quiescent H mode.

A path to a new high performance regime has been discovered in tokamaks that could improve the attractiveness of a fusion reactor. Experiments on DIII-D using a quiescent H-mode edge have navigated a valley of improved edge peeling-ballooning stability that opens up with strong plasma shaping at high density, leading to a doubling of the edge pressure over the standard H mode with edge localized modes at these parameters. The thermal energy confinement time increases as a result of both the increased pedestal height and improvements in the core transport and reduced low-k turbulence. Calculations of the pedestal height and width as a function of density using constraints imposed by peeling-ballooning and kinetic-ballooning theory are in quantitative agreement with the measurements.

Energy confinement of hydrogen and deuterium H-mode plasmas in JT-60U

Energy confinement properties for hydrogen and deuterium H-mode plasmas are investigated. The thermal energy confinement time becomes longer in deuterium by a factor of ∼1.4 than in hydrogen at a given absorbed power. When the absorbed power is fixed, the values of electron temperature Te and ion temperature Ti become explicitly higher in deuterium than in hydrogen across the whole range of minor radius while the profiles of electron density ne are almost the same. Accordingly, the effective heat diffusivity becomes relatively lower in deuterium than in hydrogen. Despite almost the same power crossing the separatrix, type-I ELM frequency for hydrogen becomes approximately double that of deuterium. When the stored energy is fixed, the spatial profiles of ne, Te and Ti become identical for both cases while higher heating power is required in the hydrogen case. The pedestal pressure is about twice as high in deuterium as that in hydrogen at a given absorbed power. The increase of the pedestal temperature is more significant for the deuterium case while the pedestal density is not changed. The poloidal beta at the H-mode pedestal is increased linearly with the increased total poloidal beta for both cases. The relation between and is almost identical regardless of the difference of the isotope species.

Confinement and ELM characteristics of H-mode plasmas in KSTAR

The latest results of confinement and edge-localized mode (ELM) characteristics of Korea Superconducting Tokamak Advanced Research (KSTAR) H-mode plasmas are reported. The estimation of fast ion contribution to the total stored energy, calculated by both the NUBEAM and ASTRA simulations, and of the effective total heating power is used to derive the thermal energy confinement time (τE,thermal), which is compared with a multi-machine database. The measured power threshold for the L–H transition (Pthr) as a function of density shows a roll-over with minimum value at . KSTAR H-mode plasmas exhibit three distinctive types of ELMs: large type-I ELMs, intermediate ELMs and a mixed (type-I and small ELM peaks) ELM regime. Power scans show that the frequency of the large ELMs increases with increasing heating power, a feature of type-I ELMs. The quality of confinement is higher for type-I and mixed ELMy H-mode (H98(y,2) ∼ 0.9–1) than for the intermediate ELM regime (H98(y,2) ∼ 0.7). Type-I ELMs have precursor-like signals from the magnetics measurement, while the other two ELM types do not. The low-field side (LFS) profile of electron temperature (Te), from the ECE measurement, and the pedestal profile of the toroidal velocity (Vt), from charge-exchange spectroscopy, show a continuous build up on the LFS during the inter-ELM period. However, the pedestal ion temperature (Ti) remains unchanged for most of the inter-ELM period until it rapidly rises in the last stage of the ELM cycle (⩾70–80%). The estimated electron pedestal collisionality for a type-I ELMy regime is . The confinement and ELM characteristics for the ELM suppression discharges by the application of an n = 1 magnetic perturbation (MP) have also been investigated for each of the identified stages during the MP application. A second L–H transition during the L-mode phase after the end of first H-mode stage occurs for some discharges when the divertor configuration is restored by the plasma control system. Characteristics of this late H-mode are compared with those for the main H-mode.

Simulation of microtearing turbulence in national spherical torus experiment

Thermal energy confinement times in National Spherical Torus Experiment (NSTX) dimensionless parameter scans increase with decreasing collisionality. While ion thermal transport is neoclassical, the source of anomalous electron thermal transport in these discharges remains unclear, leading to considerable uncertainty when extrapolating to future spherical tokamak (ST) devices at much lower collisionality. Linear gyrokinetic simulations find microtearing modes to be unstable in high collisionality discharges. First non-linear gyrokinetic simulations of microtearing turbulence in NSTX show they can yield experimental levels of transport. Magnetic flutter is responsible for almost all the transport (∼98%), perturbed field line trajectories are globally stochastic, and a test particle stochastic transport model agrees to within 25% of the simulated transport. Most significantly, microtearing transport is predicted to increase with electron collisionality, consistent with the observed NSTX confinement scaling. While this suggests microtearing modes may be the source of electron thermal transport, the predictions are also very sensitive to electron temperature gradient, indicating the scaling of the instability threshold is important. In addition, microtearing turbulence is susceptible to suppression via sheared E × B flows as experimental values of E × B shear (comparable to the linear growth rates) dramatically reduce the transport below experimental values. Refinements in numerical resolution and physics model assumptions are expected to minimize the apparent discrepancy. In cases where the predicted transport is strong, calculations suggest that a proposed polarimetry diagnostic may be sensitive to the magnetic perturbations associated with the unique structure of microtearing turbulence.

Characteristics of the first H-mode discharges in KSTAR

Typical ELMy H-mode discharges have been obtained in the KSTAR tokamak with the combined auxiliary heating of neutral beam injection (NBI) and electron cyclotron resonant heating (ECRH). The minimum external heating power required for the L–H transition is about 0.9 MW for a line-averaged density of ∼2.0 × 1019 m−3. There is a clear indication of the increase in the L–H threshold power with decreasing density for densities lower than ∼2 × 1019 m−3. The L–H transitions typically occurred shortly after the beginning of plasma current flattop (Ip = 0.6 MA) period and after the fast shaping to a highly elongated double-null divertor configuration. The maximum heating power available was marginal for the L–H transition, which is also implied by the relatively slow transition time (>10 ms) and the synchronization of the transition with large sawtooth crashes. The initial analysis of thermal energy confinement time (τE) indicates that τE is higher than the prediction of multi-machine scaling laws by 10–20%. A clear increase in electron and ion temperature in the pedestal is observed in the H-mode phase but the core temperature does not change significantly. On the other hand, the toroidal rotation velocity increased over the whole radial range in the H-mode phase. The measured ELM frequency was around 10–30 Hz for the large ELM bursts and 50–100 Hz for the smaller ones. In addition, very small and high frequency (200–300 Hz) ELMs appeared between large ELM spikes when the ECRH is added to the NBI-heated H-mode plasmas. The drop of total stored energy during a large ELM is up to 5% in most cases.

Collisionality and safety factor scalings of H-mode energy transport in the MAST spherical tokamak

A factor of 4 dimensionless collisionality scan of H-mode plasmas in MAST shows that the thermal energy confinement time scales as . Local heat transport is dominated by electrons and is consistent with the global scaling. The neutron rate is in good agreement with the ν* dependence of τE,th. The gyrokinetic code GYRO indicates that micro-tearing turbulence might explain such a trend. A factor of 1.4 dimensionless safety factor scan shows that the energy confinement time scales as . These two scalings are consistent with the dependence of energy confinement time on plasma current and magnetic field. Weaker qeng and stronger ν* dependences compared with the IPB98y2 scaling could be favourable for an ST-CTF device, in that it would allow operation at lower plasma current.