ITER Baseline Scenario plasmas were studied in DIII-D using krypton and xenon gases as a proxy for the tungsten that will be present in ITER. These impurities were chosen for having the same radiative loss rate Lz as tungsten would exhibit in the hotter ITER core. Results show that the scenario with these core radiators spans the range of impurity concentration and W radiated fraction expected for ITER, and up to 50% higher values, explored at zero injected torque, as well as 1 Nm and full co-torque injection with T similar to 3 Nm. Stationary discharges with duration >2-4 tau(R) are achieved with f(rad)>= 30% leading to a reduction in confinement of similar to 10%, and a comparison with real metal radiators in the same range of f(rad) shows that the higher Lz at the lower temperatures in these plasmas yields too pessimistic results on the survivability and performance of this scenario in ITER. Simulations of ITER power balance including W radiation show that with concentration up to three times higher than in the DIII-D plasmas the scenario can be stationary, remaining at acceptable core radiated fraction values.

Experiments in DIII-D document the ITER Baseline Scenario (IBS) at q 95 ∼ 3 and P IN / P LH ∼ 1–2, in both deuterium and hydrogen utilizing Kr and Xe as Tungsten-equivalent radiators. The power threshold for H-mode operation ( P LH ) was determined experimentally without added impurities and found to be about a factor of two higher than the scaling law. In recent IBS experiments in deuterium, intrinsic levels of metals such as Tungsten (W) or molybdenum and inconel are present that reduce the pedestal pressure by 20%–25%. A complete radiative collapse of deuterium IBS plasmas occurs at W core concentrations C W = 10 −5 . Simulations show that for core temperatures expected for ITER, the plasmas would not have a radiative collapse at C W = 1 × 10 −5 , moreover Q = 8–10 would still be achieved for C W up to 3 × 10 −5 . In contrast to deuterium, the IBS in hydrogen is not affected by intrinsic high-Z impurities, indicating that hydrogen H-modes in ITER may not inform the D-T phase with respect to W accumulation and discharge survival. Compared to deuterium, the pedestal pressure in hydrogen is ∼25% lower, with much higher ELM frequency of 150 Hz, decreasing with input power. Krypton was injected in a matrix scan of input power and impurity flow in IBS hydrogen discharges. Krypton impurity density profiles in hydrogen are similar to deuterium plasmas, but at Kr flows that are 2–3 times higher for the same input power. Krypton is transported into the core and affects the whole radius; at the highest injection rates a radiative collapse occurs at core radiation fractions of 0.3–0.35, consistent with the expected maximum W radiation fraction for ITER core plasmas. Comparing the results with previous International Tokamak Physics Activity database studies of the IBS confirms that at higher radiation fraction due to high-Z impurities, a drop in H 98 of >10% is observed. On the other hand, the results using Kr as a W-equivalent radiator indicate that metal (W) devices at lower core temperatures than ITER may provide overly pessimistic performance extrapolations to ITER for deuterium-tritium operation. The new DIII-D results support a more attractive option for the ITER Research Plan with a short hydrogen phase for system commissioning, transitioning to deuterium operations as soon as possible to provide relevant conditions for deuterium-tritium operations.

An experimental scan of the electron cyclotron waves (EC) toroidal injection angle in plasma breakdown is performed at the DIII-D tokamak. The second harmonic, extraordinary mode EC is used for the study. The dependence of ne and T-e on the EC injection angle cannot be conclusively drawn from this study due to the large error bars in the n(e) and T-e measurements. On the other hand, high T-e data points are observed in some discharges which can be explained by nonlinear heating. The D alpha emission measurement shows a clear relation between the breakdown time and the injection angle. An experimental investigation of the cause of the dependence of breakdown delay on the EC injection angle suggests that when the injection angle is oblique, the EC heating after the reflection at the inboard wall may become ineffective and cause the breakdown delay even when the EC heating before and directly upon injection remains effective. A preliminary run of the heat and transport balance code DYON indicates that the obtained dataset is suitable for a quantitative validation of EC absorption models.

JET, the world’s largest operating tokamak with unique Be/W wall and tritium handling capability, completed a Deuterium-Tritium (D-T) campaign in 2021 (Maggi et al 29th Fusion Energy Conf. ) following a decade of preparatory experiments, dedicated enhancements, technical rehearsals and training (Horton et al 2016 Fusion Eng. Des. 109–111 925). Operation with tritium raises significant technical, safety and scientific challenges not encountered in standard protium or deuterium operation. This contribution describes the tritium operational requirements, pulses and technical preparations, new operating procedures, lessons learned and details on the achieved operational availability and performance. The preparation and execution of the recent JET tritium experiments benefitted from the previous experience in 1991 (Preliminary Tritium Experiment), 1997 (DTE1 campaign) and 2003 (Trace Tritium Campaigns) and consisted of the following five phases: technical rehearsals and scenario preparation, tritium commissioning, 100% tritium campaign, D-T campaign (DTE2), tritium clean-up. Following the clean-up JET resumed normal operation and is currently undertaking a further D-T campaign (DTE3).

The path to fusion in the United States requires partnership between public and private sector. While the private sector provides the vigor to take some of the major steps necessary, there is a depth of expertise and capability in the public sector that is vital to resolving feasible approaches. As an open national user facility, DIII-D provides a crucial testbed to develop the required new technologies and approaches in relevant conditions. It has unparalleled potential to meet this challenge, thanks to its extreme flexibility and world leading diagnostics. This provides a basis to rapidly develop solutions that project to future reactors with confidence. The program has thus been redeveloped to enable public and private sector engagement and testing of new concepts. A new technology program has been launched to resolve plasma interacting technologies. With modest heating upgrades, the facility can confront the crucial "Integrated Tokamak Exhaust and Performance" gap, to resolve core, exhaust and technology solutions together. The device is also being redeveloped as a training facility, with dedicated student run time, a mentorship program, and open access to all opportunity roles, part of wider efforts to diversify and open pathways through inclusion, access, and equity. This exciting agenda is enabling scientists and technology researchers to pioneer the solutions needed for a Fusion Pilot Plant (FPP) and ITER this decade. As a national user facility, DIII D has singular potential to provide the tools, teams, and insight necessary, to do its part in moving the United States rapidly toward the commercialization of fusion energy.

The JET hybrid scenario has been developed from low plasma current carbon wall discharges to the record-breaking Deuterium-Tritium plasmas obtained in 2021 with the ITER-like Be/W wall. The development started in pure Deuterium with refinement of the plasma current, and toroidal magnetic field choices and succeeded in solving the heat load challenges arising from 37 MW of injected power in the ITER like wall environment, keeping the radiation in the edge and core controlled, avoiding MHD instabilities and reaching high neutron rates. The Deuterium hybrid plasmas have been re-run in Tritium and methods have been found to keep the radiation controlled but not at high fusion performance probably due to time constraints. For the first time this scenario has been run in Deuterium-Tritium (50:50). These plasmas were re-optimised to have a radiation-stable H-mode entry phase, good impurity control through edge Ti gradient screening and optimised performance with fusion power exceeding 10 MW for longer than three alpha particle slow down times, 8.3 MW averaged over 5 s and fusion energy of 45.8 MJ.

Achieving high neutron yields in today’s fusion research relies on high-power auxiliary heating in order to attain required core temperatures. This is usually achieved by means of high neutral beam (NB) and radio frequency (RF) power. Application of NB power is accompanied by production of fast beam ions and associated beam–target (BT) reactions. In standard JET operational conditions, deuterium (D) NBs are injected into D plasmas. The injected beams comprise D atoms at full, one-half and one-third injected energy. Typically, the full energy of the injected D beams is between 90 and 120 keV, providing 1.4–2.0 MW of heating, which is about half of the injected power. Half-energy D beams carry about one-third of the injected power and the rest of the power is carried by the third energy fraction of D beams. Under these conditions, thermal fusion reactions, i.e. those between plasma ions, and BT reactions are of the same order of magnitude. This study addresses important issues regarding the impact of density, central electron and ion temperatures and their ratio, T i (0)/ T e (0), on fusion performance, measured by the total neutron yield and BT neutron counts. NB/RF synergistic effects are discussed as well. It is demonstrated that thermal fusion gain increases linearly with normalised plasma pressure, β N , and confinement, B t τ . The BT neutrons are, however, more difficult to predict and this task in general requires numerical treatment. In this study, BT neutrons in JET’s best-performing baseline and hybrid pulses are analysed and the underlying dependencies discussed. Central fast ion densities are found to decrease with increased density and density peaking. This is attributed to poorer beam penetration at high density. The BT reactions however are unchanged and can even increase if operating at higher core temperatures. An increase in the central ion temperature and T i (0)/ T e (0) ratio leads to higher total and BT reaction rates whilst simultaneously the ratio of the BT to total neutron decreases significantly. NB/RF synergistic effects are found to have a negligible impact on total neutron rate. This can be explained by the reduced beam penetration in high-density conditions leading to lower central fast ion density.

The density peaking was studied in the ICRH versus the NBI identity plasmas and results in R/Ln=0.45 for the ICRH shot and R/Ln=0.93 for the NBI shot. The dimensionless profiles of q, ρ*, υ*, βn and Ti/Te≈1 were matched within 5% difference except in the central part of the plasma (ρtor<0.3). The difference in the curvature pinch (same q-profile) and thermo-pinch (Ti=Te) between the ICRH and NBI shots is virtually zero. The ICRH and NBI heated identity plasmas yield similar plasma parameters and performance in the core plasma (0.3<ρtor<0.8). This result is valid at 8MW of heating power level, however, it remains to be seen how this promising result will scale to larger power levels and to larger devices. The identity experiment shows that NBI fuelling increases the density peaking by a factor of 2. This number may be modified due to different rotation, fast ion content or impurities which will be assessed in future.

This article investigates the triggering of ELMs on JET by injection of frozen pellets of isotopes of Hydrogen. A method is established to determine the probability that a specific pellet triggers a particular ELM. This method allows clear distinction between pellet-ELM pairs that are very likely to represent triggering events and pairs that are very unlikely to represent such an event. Based on this, the pellet parameters that are most likely to affect the ability of pellets to trigger ELMs have been investigated. It has been found that the injection location is very important, with injection from the vertical high field side showing a much higher triggering efficiency than low field side (LFS) injection. The dependence on parameters such as pellet speed and size and the time since the last ELM is also seen to be much stronger for LFS injection. Finally, the paper illustrates how improvements to the pellet injection system by streamlining the pellet flight lines and slightly increasing the pellet size has resulted in a significantly improved ability to deliver pellets to the plasma and trigger ELMs.

The pedestal structure, edge transport and linear MHD stability have been analyzed in a series of JET with the ITER-like wall hydrogen (H) and deuterium (D) type I ELMy H-mode plasmas. The pedestal pressure is typically higher in D than in H at the same input power and gas rate, with the difference mainly due to lower density in H than in D (Maggi et al (JET Contributors) 2018 Plasma Phys. Control. Fusion 60 014045). A power balance analysis of the pedestal has shown that higher inter-ELM separatrix loss power is required in H than in D to maintain a similar pedestal top pressure. This is qualitatively consistent with a set of interpretative EDGE2D-EIRENE simulations for H and D plasmas, showing that higher edge particle and heat transport coefficients are needed in H than in D to match the experimental profiles. It has also been concluded that the difference in neutral penetration between H and D leads only to minor changes in the upstream density profiles and with trends opposite to experimental observations. This implies that neutral penetration has a minor role in setting the difference between H and D pedestals, but higher ELM and/or inter-ELM transport are likely to be the main players. The interpretative EDGE2D-EIRENE simulations, with simultaneous upstream and outer divertor target profile constraints, have indicated higher separatrix electron temperature in H than in D for a pair of discharges at low fueling gas rate and similar stored energy (which required higher input power in H than in D at the same gas rate). The isotope dependence of linear MHD pedestal stability has been found to be small, but if a higher separatrix temperature is considered in H than in D, this could lead to destabilization of peeling-ballooning modes and shrinking of the stability boundary, qualitatively consistent with the reduced pedestal confinement in H.

This paper compares the stationary ELMy H-mode baseline 50%-50% deuterium-tritium (DT) mixture discharges in the first high -power JET DT experimental campaign in 1997 (JET-DTE1) with the counterpart 100% deuterium (DD) baseline discharges which have the same engineering parameters, i.e. I-p, B-t, q(95), NBI power, plasma shape, no gas dosing, divertor structure, and the carbon wall. There is no difference in profile peaking of T-e, T-i, and n(e) between the DT and the counterpart DD baseline discharges, indicating that there was no isotopic effect on the core transport in the stationary baseline DT discharges in the 1997 JET-DTE1. The core values of T-e, T-i, and n(e) are higher in some DT discharges compared to their counterpart DD discharges, but this is attributed to the higher pedestal values, rather than any improvement in the core transport. The interpretive TRANSP simulations also show that the local heat diffusivity is not consistently different between the DT and the counterpart DD baseline discharges. The baseline discharges in the 1997 JET-DTE1 are also compared to the latest high-power ELMy H-mode baseline DD discharges with an ITER-like wall (ILW) in 2016. Despite the similar effective collisionality and ion heat deposition in the core, it was observed that T-i/T-e is consistently close to unity in the 1997 JET-DTE1 discharges, while the 2016 JET baseline discharges have high T-i, exceeding T-e, which enabled the highest fusion performance in the ITER-Like Wall. The high rotation frequency was the key factor in increasing T-i/T-e in the 2016 JET baseline discharges, and it is also the main difference compared to the stationary baseline discharges in the 1997 JET-DTE1. Based on this, it is prospected that higher rotation frequency is the key factor to achieving high fusion power in the stationary baseline discharges in the 2020 JET-DTE2, and the plasma operation with the low gas dosing and increased torque available in the present NBI system would enable such a high rotation frequency.

Joint European Torus (JET) plasma initiations that form a significant quantity of runaway electrons have been studied. It is shown that there is no direct relationship between the prefill pressure and breakdown electric field and signs of runaway electrons during the plasma initiation. Runaway electron generation is determined by the electric field and density development at and after burn-through. A clear criterion of density and electric field at one given point in time, which would ensure the avoidance of runaway electron generation, cannot be determined, because the timescales for the formation of runaway electrons and for the dynamics of the density differ significantly. Moreover, the formation process can be reversed, reducing the influence of runaway electrons on the discharge. Ensuring a high enough density will reduce the likelihood that runaway electron discharges are formed. It is also found that at JET the electric field often exceeds the critical electric field during the early stages of the current ramp-up phase, even when no signs of runaway electrons are present. Expected runaway current dynamics have been analysed using the discharge circuit equation. The comparison of the expected runaway electron current dynamics shows it to be significantly slower compared to theoretical expectations in the presence of a hot and dense thermal background plasma. This could be explained by an enhanced critical electric field and/or a reduced confinement of runaway electrons. The latter is shown to be affected by bursts of magnetohydrodynamic activity that are characteristic during the current ramp-up. The development of discharges in which the current is fully carried by runaway electrons happens on a slow timescale of several seconds, limited by the available flux. Such timescales are sufficient for improved active control of these events, avoiding runaway currents at plasma initiation exceeding values at which they could damage in-vessel components. The results provide insight into the improvement and interpretation of self-consistent modelling of runaway electron generation during the start-up of International Thermonuclear Experimental Reactor discharges.

The present paper offers an overview of the potential of ion cyclotron resonance heating (ICRH) or radio frequency heating for the DEMO machine. It is found that various suitable heating schemes are available. Similar to ITER and in view of the limited bandwidth of about 10 MHz that can be achieved to ensure optimal functioning of the launcher, it is proposed to make core second harmonic tritium heating the key ion heating scheme, assisted by fundamental cyclotron heating He-3 in the early phase of the discharge; for the present design of DEMO-with a static magnetic field strength of B-o = 5.855 T-that places the T and 3He layers in the core for f = 60 MHz and suggests centering the bandwidth around that main operating frequency. In line with earlier studies for hot, dense plasmas in large-size magnetic confinement machines, it is shown that good single pass absorption is achieved but that the size as well as the operating density and temperature of the machine cause the electrons to absorb a non-negligible fraction of the power away from the core when core ion heating is aimed at. Current drive and alternative heating options are briefly discussed and a dedicated computation is done for the traveling wave antenna, proposed for DEMO in view of its compatibility with substantial antenna-plasma distances. The various tasks that ICRH can fulfill are briefly listed. Finally, the impact of transport and the sensitivity of the obtained results to changes in the machine parameters is commented on.

This paper presents results of JET ITER-like wall L-mode experiments in hydrogen and deuterium (D) plasmas, dedicated to the study of the isotope dependence of ion heat transport by determination of the ion critical gradient and stiffness by varying the ion cyclotron resonance heating power deposition. When no strong role of fast ions in the plasma core is expected, the main difference between the two isotope plasmas is determined by the plasma edge and the core behavior is consistent with a gyro-Bohm scaling. When the heating power (and the fast ion pressure) is increased, in addition to the difference in the edge region, also the plasma core shows substantial changes. The stabilization of ion heat transport by fast ions, clearly visible in D plasmas, appears to be weaker in H plasmas, resulting in a higher ion heat flux in H with apparent anti-gyro-Bohm mass scaling. The difference is found to be caused by the different fast ion pressure between H and D plasmas, related to the heating power settings and to the different fast ion slowing down time, and is completely accounted for in non-linear gyrokinetic simulations. The application of the TGLF quasi-linear model to this set of data is also discussed.

The need for predictive capabilities greater than 95% with very limited false alarms are demanding requirements for reliable disruption prediction systems in tokamaks such as JET or, in the near future, ITER.The prediction of an upcoming disruption must be provided sufficiently in advance in order to apply effective disruption avoidance or mitigation actions to prevent the machine from being damaged.In this paper, following the typical machine learning workflow, a generative topographic mapping (GTM) of the operational space of JET has been built using a set of disrupted and regularly terminated discharges.In order to build the predictive model, a suitable set of dimensionless, machine-independent, physics-based features have been synthesized, which make use of 1D plasma profile information, rather than simple zero-D time series.The use of such predicting features, together with the power of the GTM in fitting the model to the data, obtains, in an unsupervised way, a 2D map of the multi-dimensional parameter space of JET, where it is possible to identify a boundary separating the region free from disruption from the disruption region.In addition to helping in operational boundaries studies, the GTM map can also be used for disruption prediction exploiting the potential of the developed GTM toolbox to monitor the discharge dynamics.Following the trajectory of a discharge on the map throughout the different regions, an alarm is triggered depending on the disruption risk of these regions.The proposed approach to predict disruptions has been evaluated on a training and an independent test set and achieves very good performance with only one tardive detection and a limited number of false detections.The warning times are suitable for avoidance purposes and, more important, the detections are consistent with physical causes and mechanisms that destabilize the plasma leading to disruptions.

In tokamak H-mode plasmas, the level of energy and particle transport at the plasma edge is reduced and a steep pressure gradient is formed, giving rise to an edge pressure pedestal, which positively affects the global energy confinement. A positive isotope mass scaling of the thermal energy confinement time in H-mode plasmas has been observed in several tokamaks, however, this favourable isotope dependence has not yet been fully understood theoretically. This thesis examines the pedestal structure, edge transport, linear MHD stability and inter-ELM edge current evolution in a series of JET-ILW Hydrogen (H) and Deuterium (D) type I ELMy H-mode plasmas with the aim to better understand the isotope dependence of the pedestal and its contribution to the favourable isotope scaling. Simulations of the inter-ELM edge current evolution showed that current diffusion contributes little to the time evolution of the total edge current prior to the ELM crash. Therefore, current diffusion does not explain why JET-ILW type I ELMy pedestals at high gas rate and moderate to high plasma beta are found to be stable to Peeling-Ballooning modes. The pedestal pressure is typically higher in D than in H at the same input power and gas rate, with the difference mainly due to lower density in H than in D. Analysis of the pedestal structure and power balance, and results of interpretative 2D edge transport simulations with EDGE2D-EIRENE indicate that the difference in neutral penetration between H and D leads only to minor changes in the pedestal density and temperature profiles, and differences in heat and particle transport must also play a role in the favourable isotope scaling of the pedestal. The effect of the isotope mass on linear MHD pedestal stability is small, but an indirect isotope dependence through the separatrix temperature is qualitatively consistent with the reduced pedestal confinement in H and could play a role in JET-ILW H-mode plasmas at low gas rate.

The First Mirror Test for ITER has been carried out in JET with mirrors exposed during: (i) the third ILW campaign (ILW-3, 2015-2016, 23.33 h plasma) and (ii) all three campaigns, i.e. ILW-1 to ILW-3: 2011-2016, 63,52 h in total. All mirrors from main chamber wall show no significant changes of the total reflectivity from the initial value and the diffuse reflectivity does not exceed 3% in the spectral range above 500 nm. The modified layer on surface has very small amount of impurities such as D, Be, C, N, O and Ni. All mirrors from the divertor (inner, outer, base under the bulk W tile) lost reflectivity by 20-80% due to the beryllium-rich deposition also containing D, C, N, O, Ni and W. In the inner divertor N reaches 5 x 10(17) cm(-2), W is up to 4.3 x 10(17) cm(-2), while the content of Ni is the greatest in the outer divertor: 3.8 x 10(17) cm(-2). Oxygen-18 used as the tracer in experiments at the end of ILW-3 has been detected at the level of 1.1 x 10(16) cm(-2). The thickness of deposited layer is in the range of 90 nm to 900 nm. The layer growth rate in the base (2.7 pm s(-1)) and inner divertor is proportional to the exposure time when a single campaign and all three are compared. In a few cases, on mirrors located at the cassette mouth, flaking of deposits and erosion occurred.

We report results of benchmarking of core particle transport simulations by a collection of codes widely used in transport modelling of tokamak plasmas. Our analysis includes formulation of transport equations, difference between electron and ion solvers, comparison of modules of the pellet and edge gas fuelling on the ITER baseline scenario. During the first phase of benchmarking we address the particle transport effects in the stationary phase. Firstly, simulations are performed with identical sources, sinks, transport coefficients, and boundary conditions prescribed in the flattop H-mode phase. The transformation of ion particle transport equations is introduced so to directly compare their results to electron transport solvers. Secondly, the pellet fuelling models are benchmarked in various conditions to evaluate the dependency of the pellet deposition on the pellet volume, injection side, pedestal, and separatrix parameters. Thirdly, edge gas fuelling is benchmarked to assess sensitivities of source profile predictions to uncertainties in plasma conditions and detailed model assumptions. At the second phase, we address particle transport effects in the time-evolving plasma including the current ramp-up to the ramp-down phase. The ion and the electron solvers are benchmarked together. Differences between the simulation results of the solvers are investigated in terms of equilibrium, grid resolution, radial coordinate, radial grid distribution, and plasma volume evolution term. We found that the selection of the radial coordinate can yield prominent differences between the solvers mainly due to differences in the edge grid distribution. The simulations reveal that electron and ion solvers predict noticeably different density peaking for the same diffusion and pinch velocity while with the peaked profile of helium, expected in fusion reactors. The fuelling benchmarking shows that gas puffing is not efficient for core fuelling in H-modes and density control should be done by the high field side pellet injection in contrast to present machines.

The analysis of the polarimetry measurements has the aim of validating models (De Marco and Segre 1972 Plasma Phys. 14 245), with a careful attention to the clarification of their limits of application. In this paper a new approximation method is introduced, the so-called special constant Omega direction (SCOD), which gives an analytical solution to the polarimetry exact Stokes model equations. The available approximate solutions (including SCOD) of the polarimetry propagation equations are presented, compared and their application limits determined, using a reference tokamak configuration, which is a simplified equilibrium for a circular tokamak. The SCOD approximation is compared successfully to the Stokes model in the context also of equilibria evaluated for two JET discharges. The approximation methods are analytical or very simple mathematical expressions which can also be used in equilibrium codes for their optimization.

This paper outlines an approach towards improved rigour in tokamak turbulence transport model validation within integrated modelling. Gaussian process regression (GPR) techniques were applied for profile fitting during the preparation of integrated modelling simulations allowing for rigourous sensitivity tests of prescribed initial and boundary conditions as both fit and derivative uncertainties are provided. This was demonstrated by a JETTO integrated modelling simulation of the JET ITER-like-wall H-mode baseline discharge #92436 with the QuaLiKiz quasilinear turbulent transport model, which is the subject of extrapolation towards a deuterium-tritium plasma. The simulation simultaneously evaluates the time evolution of heat, particle, and momentum fluxes over similar to 10 confinement times, with a simulation boundary condition at rho(tor) = 0.85. Routine inclusion of momentum transport prediction in multi-channel flux-driven transport modelling is not standard and is facilitated here by recent developments within the QuaLiKiz model. Excellent agreement was achieved between the fitted and simulated profiles for n(e), T-e, T-i, and Omega(tor) within 2 sigma, but the simulation underpredicts the mid-radius Ti and overpredicts the core n(e) and T-e profiles for this discharge. Despite this, it was shown that this approach is capable of deriving reasonable inputs, including derivative quantities, to tokamak models from experimental data. Furthermore, multiple figures-of-merit were defined to quantitatively assess the agreement of integrated modelling predictions to experimental data within the GPR profile fitting framework.