Recent Publications

Effect of pedestal current on the density window for ELM suppression using n = 4 RMP in EAST

Xuemin Wu, Youwen Sun, Qun Ma, Shuai Gu, Manni Jia, Yueqiang Liu, Yifeng Wang, Cheng Ye, Pengcheng Xie, Alberto Loarte, et al.

Existence of operational window in both edge safety factor and line averaged plasma density for suppression of ELMs using n=4 Resonant Magnetic Perturbations in low input torque plasmas has been observed in EAST experiment, in which q95 and plasma normalized beta (βN) close to that required in ITER high-Q operation. Here, n is toroidal mode number of the magnetic perturbation. In contrast to previous reports from other tokamaks, there is not only an upper density limit but also a lower one for accessing ELM suppression. Modelling results using the MARS-F code show that the RMP with linear plasma response has a peak at an intermediate density and decays as the density increases or decreases, which results in a minimal RMP field penetration threshold at the intermediate density. In this experiment, the observed lower density limit operationally manifests a sensitivity of the q-profile: different densities alter the edge current profile, which change the alignment of the eigenmode structure with the RMP coil configuration, causing a reduction of the resonant field in both low- and high-density cases, and hence making field penetration more difficult. The modelled window of the strongest resonant plasma response in terms of [⟨ne⟩, q95] agrees well with the observed ELM suppression in EAST. Peeling-ballooning modes stability analysis using the ELITE code shows that plasmas gradually approach peeling instability boundary caused by increase of edge bootstrap current as the plasma density decreases, which is consistent with the observation that ELMs come back again in lower density plasmas for fixed q95. These results indicate that linear modelling with full toroidal geometry can well predict the optimized RMP configuration for ELM suppression and reveal the important roles of pedestal plasma current, which need to be carefully considered in the application of high n RMPs for ELM suppression in future ITER.

BLUEMIRA: a modular, open-source framework for designing tokamak fusion reactors

Matti Coleman, James Cook, Fabrizio Franza, Ivan Maione, Simon Mcintosh, Hudson Baker, Alexander Ian Blair, Shail Desai, Oliver Funk, Georgina Graham, et al.

As efforts across the world to deliver fusion power enter their respective conceptual design stages, many begin to encounter the difficulties of generating complete designs, evaluating them, performing trade-off studies, and comparing different alternatives. Many organisations successfully use fusion reactor systems codes with low-fidelity formulations of the critical physics and engineering aspects to inform the initial design stage and steer them towards a region of the design space they consider attractive, but struggle to explore the design space in more depth. This work is an overview of the open-source tokamak reactor design framework, BLUEMIRA, built from the BLUEPRINT [1, 2] and MIRA [3, 4] codes, which goes beyond the “classical” systems code paradigm and enables the user to parameterise their reactor designs at higher fidelity. Reactor design workflows ranging from 0-D “radial build” design, equilibrium design, to 3-D CAD and automated neutronics analyses can be run within minutes. We present the functionality implemented in the BLUEMIRA code and demonstrate its application to the conceptual design of a conventional aspect ratio fusion reactor, performing an indicative design study over aspect ratio and number of toroidal field coils.

Electromagnetic analysis of beams for the ITER ECH Upper Launcher

P. Platania, A. Simonetto, J.M. Arroyo, E. Carbonell, N. Casal, M. Choe, D. Farina, G. Garre, M. Henderson, M. Jimenez, et al.

3D hot tail runaway electron generation modelling for ITER

Louis Puel, Eric Nardon, F J Artola, Di Hu

Runaway electron (RE) generation represents a major concern for future tokamaks such as ITER, where high plasma temperatures and large plasma currents significantly enhance the risk of RE formation. In particular, the hot tail mechanism, triggered during the thermal quench (TQ), occurs when a rapid drop in plasma temperature prevents high-energy electrons from thermalizing, allowing them to become REs under the accelerating effect of a strong toroidal electric field. In ITER, this mechanism is expected to be the dominant primary generation before avalanching in the case of an unmitigated/poorly mitigated disruption. The recently developed hot tail tracker framework in JOREK is used here to perform the first 3D estimates of hot tail generation during disruptions of an ITER 15 MA D–T H-mode scenario mitigated by Shattered Pellet Injection (SPI). Two situations are considered, with and without pre-disruption thermal energy degradation. The non-degraded case, more pessimistic, results in strong RE generation driven by weak stochastic losses limited by the trapped population, strong helical cooling, and the migration of hot electrons from the core toward regions of high parallel electric field. In contrast, the degraded case shows successful mitigation within the assumptions of this simulation, with suppression of the core hot tail generation and the removal of the edge seed through stochastic transport. The difference between a critical and a non critical scenario appears to be linked to localized overdensities of injected material produced by magneto-hydrodynamic (MHD) activity, which, counter-intuitively, significantly enhance local generation. This phenomenon is confirmed by a 0D two-fluid model of the bulk plasma during the TQ, coupled with a new 0D hot tail estimator based on backward tracking of a single particle in velocity space.

A simplified model analysis of runaway electron behavior following magnetic stochasticity and healing in vertically unstable plasmas during tokamak disruptions

Jose Ramon Martin Solis, Jose Angel Mier, F J Artola, Alberto Loarte

A simplified model analysis is used for the investigation of the survival of runaway electrons following the break-up of the magnetic field lines and the reformation of the flux surfaces during the current-quench phase of tokamak disruptions in vertically unstable plasmas. It is found that even low levels of the runaway current (~ a few tens of kAs) when the plasma touches the wall might lead to runaway damage due to the strong runaway avalanche during scraping-off. Low enough runaway deconfinement times (τd < 1 ms) in a sufficiently long stochastic phase (τ/τd > 5) would be required to avoid a large energy transfer to the runaway electrons and damage of the plasma facing components for low temperatures (a few eVs) of the residual ohmic plasma. The role played by the primary runaway generation mechanisms in regenerating the runaway beam after the magnetic surfaces have healed is also discussed.

Investigation of high-Qfus L-mode plasma operation sustained by elevated pellet fuelling in ITER

Jie Zhang, Florian Koechl, Alexei R Polevoi, Clarisse Bourdelle, Sunhee Kim, Alberto Loarte, Simon D Pinches, Ge Zhuang

The enhanced confinement of tokamak plasmas (H-mode) makes it a preferred regime for achieving fusion power production goals in future devices such as ITER. Nevertheless, low confinement mode (L-mode) remains worthy to investigate in reactor relevant conditions, primarily due to no/reduced requirements for ELM and divertor heat load control. In this regard, this study aims at exploring a new potential approach to maximise the achievable fusion gain Qfus. This approach attempts to increase the core density with enhanced pellet fuelling and then investigates the feasibility of high Qfus L-mode operation in ITER. The JINTRAC integrated modelling suite has been employed for core-edge transport and source modelling, using the HPI2 module for pellet fuelling. In some of the scenarios considered, the core density reaches up to ∼185% of the Greenwald density, nGW, with edge densities approaching nGW, motivated by recent re-evaluations of the density limit that suggest a power-dependent threshold. We compare core transport modelling results obtained by applying the semi-empirical Bohm-gyro-Bohm (BgB) or the quasi-linear gyrokinetic TGLF-SAT2 anomalous transport models, with interpretive vs. predictive impurity transport modelling, and pellet fuelling describing continuous vs. discrete particle sources. The core plasma confinement of high-density L-mode operation in ITER predicted by the TGLF-SAT2 model is significantly better than that predicted by the BgB model, resulting in a significantly improved Qfus. Fusion performance metrics, including Pfus and Qfus, exhibit only minor changes when switching from the interpretive impurity model to the predictive SANCO model, and/or from the continuous ad-hoc pellet model to the discrete HPI2 pellet model. The highest Qfus value predicted in the ITER high-density L-mode simulations is ~ 4, with indications that further improvement may be limited by increased transport associated with electro-magnetic turbulence at elevated plasma beta. This integrated modelling prediction demonstrates the potential of improved Qfus L-mode operation in ITER and future fusion devices, while exploring its boundary.

Coordinate-invariant flux-surface Fourier analysis in tokamaks

Matthew Pharr, Evan Bursch, Nikolas Logan, Priyansh Lunia, Jong-Kyu Park, Carlos Paz-Soldan

The Fourier spectra of resonant quantities in tokamaks depend on the choice of magnetic coordinates, and an area weighting of the Fourier integrand preserves the resonant coefficients on rational surfaces. That result constrains only the resonant interior; the coordinate dependence of the external Fourier spectrum, which determines the coupling to Resonant Magnetic Perturbation (RMP) coils and error-field penetration, was left untreated. This paper shows that pairing a square-root-area weighted vacuum field perturbation with a full-area-weighted resonant field yields a coupling matrix C whose singular values are invariant under coordinate transformations and whose right singular vectors reconstruct to a consistent real-space field pattern across coordinate systems, completing the coordinate-invariance picture for the plasma-3D-field coupling paradigm. GPEC calculations confirm the analytic result and show that improperly weighted coupling matrices can produce dominant modes whose overlap with the vacuum field perturbation differs by a factor of $2--3$ between coordinate systems for strongly shaped, low aspect ratio equilibria, with the discrepancy growing with inverse aspect ratio. The same coordinate dependence afflicts alternative formulations such as the three-mode metric or zeroing the $q=2$ resonant field without proper weighting. The result applies to any tool computing Fourier spectra of resonant or external quantities on flux surfaces.

Runaway electron generation in ITER mitigated disruptions with improved physics models

Lorenzo Votta, F J Artola, Eric Nardon, Oskar Vallhagen, Mathias Hoppe

We assess runaway-electron (RE) generation in ITER disruptions mitigated by shattered pellet injection (SPI) using improved physics modelling in the 1D disruption simulation framework DREAM. To this end, we extend DREAM with four ITER-relevant physics models: (i) a reduced model for RE scrape-off associated with the vertical plasma motion, (ii) a semi-analytical plasmoid-drift model for material deposition, (iii) an adaptive hyper-resistive transport model to suppress unphysical thin-current channels during the current quench (CQ), and (iv) an updated Compton RE generation seed calculated for the new ITER tungsten first-wall design. We simulate full-current 15 MA L-mode (H26, non-nuclear) and H-mode (DTHmode24, nuclear) scenarios, and an intermediate-current 7.5 MA H-mode non-nuclear case, from realistic ITER inputs. Within the adopted reduced-model framework, complete avoidance of a multi-MA RE beam is found to require a long pre-thermal quench (TQ) duration to thermalize the hot-tail electrons, high deuterium assimilation with limited neon, and a representative seed current comparable to a single RE in ITER. As previously found with lower fidelity setups [Vallhagen et al, Nucl. Fusion 64 (2024)], these conditions are met by staggered or low-Ne injections in H26, but are typically violated in DT H-mode when nuclear seeds are present. In addition to analyzing the effect of the new models, we investigate the role of the current spike associated with the TQ and the importance of radial transport of runaways in the CQ. After incorporating these additional physical effects into a comprehensive disruption model and analyzing their impact, we present a representative ITER DT H-mode SPI scenario which , within the adopted reduced-model framework, yields a substantially mitigated representative RE current. This case illustrates a possible route toward reduced RE-current levels in ITER DT operation, although its quantitative tolerability remains dependent on scenario-specific impact and termination assumptions.

Innovative lasers and polychromators for multi-wavelength real-time Thomson scattering in magnetic-confinement fusion devices

Eugene E. Mukhin, Sergey Yu. Tolstyakov, Gleb S. Kurskiev, Nikita V. Ermakov, Nikita S. Zhiltsov, Philip Ph. Forsh, Ekaterina E. Tkachenko, Valery A. Solovey, Sergey E. Aleksandrov, Andrey V. Nikolaev, et al.

Tungsten limiter start-up experiments on ASDEX Upgrade and WEST in different boronization states in support of ITER

Joerg Hobirk, Richard A Pitts, Pierre Manas, Clemente Angioni, Matthias Bernert, Dominik Brida, Guido Ciraolo, Laurent Colas, Corinne Desgranges, Ralph Dux, et al.

Understanding the performance of limiter plasmas in the ITER start-up phase is important for the whole pulse and a possible challenge if performed on tungsten, as will now be the case following the switch to a W first wall in the new 2024 ITER Baseline. Experiments were performed on ASDEX Upgrade and WEST to characterise limiter plasmas using boronizations with different degree of boron surface coverage and toroidal asymmetries. Non-boronized start-up is shown to be slow and laborious, and, on both machines, was aborted in favour of a non-homogeneous boronization, performed also in support of the ITER re- baseline to study the impact of spatially non-uniform boron coatings. This allows normal start- up and short (few 100 ms, ITER will run ≈ 10s long limiter phases) limiter phases can be run without problems. Even with a full boronization, the limiter can de-condition and long limiter plasmas of several seconds suffer from high densities and radiation. Short limiter plasmas for plasma current ramp-up remain possible. The conditioning effect on limiter plasmas is documented, but also how the start-up is affected by an ageing boronization.

Direct comparison of 3D non-linear JOREK simulations of shattered pellet injection with ASDEX Upgrade experiments

Weikang Tang, Matthias Hoelzl, Paul Heinrich, Di Hu, F J Artola, Pascal de Marne, Mathias Dibon, Mike G Dunne, Ondřej Ficker, Peter Halldestam, et al.

Shattered pellet injection (SPI) as primary mitigation method for major disruptions in ITER has a large parameter space available for optimization including the total amount of injected material, the size of the individual pellet fragments, the material composition, and the timing of multiple injections. This flexibility needs to be exploited to simultaneously minimize thermal heat loads, electromagnetic vessel forces, and formation of relativistic electrons and their impacts on plasma facing components. In this article, we apply 3D non-linear magnetohydrodynamic modelling to SPI experiments in the ASDEX Upgrade tokamak, going beyond our previous work [Tang et al Nucl. Fusion 65 116003 (2025)] by resolving some discrepancies between simulations and experiment and carrying out direct qualitative and quantitative comparisons to experimental measurements. The key element that enables the transition is the incorporation of the parallel heat-flux limit, which is done here in a simplified form. The work increases the confidence of reproducing key processes of disruption mitigation in direct 3D non-linear simulations in view of future predictive studies for ITER.

Evaluating the structural reliability of piping systems for ITER Port Cell 16: Design challenges and optimization in Test Blanket Systems

E. García-Blanco, F. Tata-Ducru, D. Guerra, J.P. Martins, F. Fantini, A. Portone, I. Ricapito, R. Rotella, J. Van der Laan

Validation of Integrated Modelling of KSTAR Plasmas with Carbon Divertor using JINTRAC and Prediction for Tungsten Divertor

Beomsu Kim, Sun Hee Kim, Florian Koechl, Francis J Casson, Romain Futtersack, Chan-Young Lee, SeulChan Hong, JAEGON LEE, Boseong Kim, Yong-Su Na

The JINTRAC integrated modelling suite has been adapted for the KSTAR tokamak, with its primary goal being the establishment of a core-edge integrated modelling framework for predictive analysis of KSTAR plasmas. This framework was validated by benchmarking JINTRAC's physics modules against both established codes and experimental data from a KSTAR carbon (C) divertor discharge (#25458), demonstrating high fidelity. Based on the validation, predictive core-edge coupled simulations were performed to investigate the effects of changing the divertor material from C to tungsten (W). The simulations predicted a significant increase in core radiation and a corresponding decrease in plasma temperatures, consistent with the behavior of high-Z materials. Specifically, the high tungsten influx causes radiative cooling in the pedestal and core regions, leading to a reduction in edge bootstrap current and a compensatory increase in core inductive current. This redistribution of current density drops the central safety factor (q 0 ) below unity, triggering sawtooth instability which serves to suppress core impurity accumulation. Sensitivity analysis indicated that while toroidal rotation exhibits a non-monotonic correlation with tungsten profile peaking, the total tungsten inventory and radiated power (P rad ) increased with higher rotation. Additionally, increasing the separatrix electron density effectively suppressed the tungsten source at the divertor target, yet the reduction in core tungsten inventory was limited compared to the extent of source reduction. These findings imply that achieving high-performance operation requires a comprehensive control strategy that minimizes the absolute impurity influx across the separatrix through combined source suppression and transport control.

A statistical study on as-built deviation of ITER blanket shield block interfaces

Ji-Young Jeong, Sa-Woong Kim, Jun-Sung Chang, Tae-Hwan Kwon, Duck-Hoi Kim

Evolution of SPI-induced disruptions in ASDEX Upgrade

Paul Heinrich, Gergely Papp, Stefan Jachmich, F J Artola, Matthias Bernert, Pascal de Marne, Mathias Dibon, Ralph Dux, Thomas Eberl, Ondřej Ficker, et al.

Disruptions are a major concern for future fusion reactors based on the tokamak principle. To ensure machine protection, the thermal loads and vessel forces that arise during disruptions have to be mitigated reliably. For the ITER disruption mitigation system (DMS), the shattered pellet injection (SPI) technology has been selected. It can provide a prompt delivery of the injection material into the plasma core, with the mitigation efficiency depending on fragment size and velocity. A highly flexible SPI system was built and installed at the tokamak ASDEX Upgrade (AUG) to aid the finalization process of the ITER DMS and provide crucial input for modeling. The SPI-induced disruptions in the 2022 AUG experiments follow a typical chain of events, which are discussed in this paper: The first light (FL), main fragment arrival (MFA), plasma movement event (PME), MARFE, thermal quench (TQ)/plasma current spike (IP-spike), current quench (CQ), and vertical displacement event (VDE) phase. Depending on the injection parameters, these phases may vary significantly or some might not be present at all. In this paper, we will focus on the characterization of these disruption phases and figures of merit for the mitigation efficiency, depending on the SPI configuration. With increasing amount of assimilated neon in the plasma - primarily influenced by the neon content in the pellet but also the shattering parameters - the disruptions exhibit different behaviors. This disruption evolution seems to be a continuous process, with the most prominent feature being the changing disruption time scales and plasma current time trace shape during the CQ from convex (poorly or unmitigated) → concave (well mitigated/radiation dominated). Depending on the injection, pre-TQ durations between 15 - 0.5 ms and early CQ durations (Δt CQ 100 → 80 ) between 13.3 - 8.2 ms had been achieved at AUG.

Beam optics analysis by visible cameras applied to the multi-beamlet configuration in the ITER NBI prototype source

Giulia Emma, Margherita Ugoletti, Matteo Agostini, Riccardo Agnello, Riccardo Casagrande, Isabella Mario, Roberto Pasqualotto, Antonio Pimazzoni, Carlo Poggi, Basile Pouradier-Duteil, et al.

The production of high energy neutral hydrogen/deuterium beams (0.87 MeV/1 MeV) with more than 90% uniformity and extremely low divergence (< 7 mrad) is highly challenging for the realization of the ITER Neutral Beam Injection system. The ITER prototype negative ion source, SPIDER, is in operation at the Neutral Beam Test Facility in Padua and it is equipped with a full set of diagnostics to study and investigate the beam properties. Among them, a system of 15 2D visible cameras, surrounding the vacuum vessel, is employed to measure the beam uniformity and divergence by detecting the light emitted from the beam-background gas interaction. In this paper, the beam divergence measured by these cameras is evaluated for the first time in the multi-beamlet configuration explored during the most recent SPIDER operation. The analysis is performed using the top-view and the side-view cameras, studying both the horizontal and vertical beam profiles through a 1D Gaussian fitting procedure, to estimate the width of the beamlet column or row aligned with the camera's lines of sight. The divergence is then retrieved from the linear fit of the width measured at different positions along the beam propagation direction. The method is here presented and applied to the experimental data acquired with an open beam segment. The dependence of the divergence on the main source and accelerator parameters, namely the RF power, the source filling pressure, and the voltage of the accelerator grids, is investigated. The results are then compared with measurements of the divergence given by the other beam diagnostics available in SPIDER, showing good agreement and validating the use of the visible camera system for the beam optics characterisation in the multi-beamlet operation.

Alpha particle velocity- and orbit-space sensitivity of gamma-ray spectroscopy diagnostics based on the 10 B(α,pγ) 13 C reaction

Massimo Nocente, Alessandro Ciurlino, Andrea Valentini, Henrik Järleblad, Yevgen Kazakov, Vasily Kiptily, Mads Rud Larsen, Bruno Coriton, Gabriele Croci, Andrea Dal Molin, et al.

A key challenge on the path towards burning plasmas is the measurement of the alpha particle phase space. Among the few experimental options available, gamma-ray spectroscopy is a key method. Previous demonstrations at JET were based on reactions between alpha particles and beryllium as the target impurity, but this is no longer applicable, neither at ITER with a tungsten-based first wall, nor in other forthcoming burning plasma experiments. On the other hand, owing to the use of boron injection for plasma operations in tungsten devices, gamma-ray reactions between the alpha particles and boron impurities have been proposed as an alternative. This work presents the first numerical evaluation of the velocity-space and orbit-space sensitivity of boron-based gamma-ray spectroscopy to the alpha particle phase space, focusing on the 10 B(α,pγ) 13 C reaction. Depending on the emission peak, we find that measurements can be separately sensitive to selected super- and sub-alfvenic velocities, with an enhanced sensitivity towards particles on trapped and potato orbits, for the case of radial lines of sight. Information on co- and counter-going alphas may be accessed only by more refined spectral shape analysis or if a tangential line of sight is deployed. The results are relevant to establish a bridge between the indirect gamma-ray spectroscopy measurements and the alpha particle phase space. They furthermore provide a quantitative input to determine which wave-alpha particle resonant interactions may (or may not) be measurable with boron based gamma-ray spectroscopy, depending on the observation geometry.

Kinetic equilibrium prediction at TCV using RAPTOR and FBT

C.E. Contré, A. Merle, O. Sauter, S. Van Mulders, R. Coosemans, G. Durr-Legoupil-Nicoud, F. Felici, O. Février, C. Heiss, B. Labit, et al.

We present results from a new Kinetic-Equilibrium Prediction workflow and shot preparation for full Tokamak à Configuration Variable (TCV) discharges, by coupling predict-first RAPTOR transport simulations with FBT inverse equilibrium calculations. RAPTOR is a 1.5D transport code which has been extensively used for plasma shot optimization and real-time modeling. We show that rapid pre-shot simulations can be performed directly using information from the pulse schedule across a wide range of plasma shapes and scenarios, given an estimate of the confinement quality factor H 98 ( y , 2 ) and line-averaged density. The resulting p ′ and T T ′ profiles are then provided to the pre-shot equilibrium computation performed by FBT—a static free-boundary solver routinely used at TCV—achieving convergence between the two codes in a few iterations. Finally, we show that this coupling, when integrated into the TCV shot preparation, improves the evaluation of the coil currents needed to match the target plasma shape; in particular providing an accurate estimate of critical quantities such as the internal inductance l i and normalized pressure β N , giving more realistic information to tokamak operators about the expected pulse behavior and enabling them to adjust the plan correspondingly.

Impact of tungsten plasma facing components on H-mode operational space in EAST in support of ITER new baseline

Manni Jia, Alberto Loarte, Youwen Sun, Qingquan Yang, Hua Yang, Shuai Gu, Ling Zhang, Wenmin Zhang, Tianqi Jia, Hui Sheng, et al.

Recent experiments in EAST have demonstrated the compatibility of tungsten (W) as main wall plasma facing material with high confinement H-mode with low or no boron coverage. The experiments were conducted in plasmas with q 95 ≈ 6.0, which allowed access to both type-I and type-II ELMy H-modes in EAST with a W wall and low normalized input torque similar to ITER. Central electron cyclotron (EC), neutral beam injection (NBI) and lower hybrid waves (LHW) were applied as auxiliary heating in a range of total power injected into the plasma from 3 MW to 5 MW. Small amplitude high frequency type-II ELMs allow maintaining good H-mode energy confinement even when the distance between the separatrix and the main W limiter is as low as 4 cm. The normalized H-mode energy confinement can reach H 98 factors up to 1.1 for both EC+NBI and EC+LHW power combination in the type-II ELMy H-mode regime. In addition, for these type-II ELMy H-mode conditions, nitrogen puffing from divertor region has been used to achieve partial detachment without significant impact on H 98 nor on the core W concentration. On the contrary, EAST operation in the type-I ELMy H-mode is strongly affected by the main wall W source. The use of n = 2 resonant magnetic perturbations (RMPs) to achieve type-I ELM suppression in EAST reduces the core W level but at the cost of about 10% reduction on energy confinement. These results complement existing W-wall findings, providing a preliminary foundation from EAST for evaluating the impact of the W first wall in ITER, as proposed in the new ITER baseline, and highlight that the achievement of sufficient ELM control levels with low deterioration of energy confinement is key to minimize this impact.

Features of fusion power measurements in next-generation magnetic plasma confinement experiments

V. Krasilnikov, T. Kormilitsyn, D. Fridrikhsen, E. Afanasenko, A. Kovalev, Y. Kashchuk, L. Bertalot, A. Dzhurik, S. Obudovsky, B. Coriton, et al.

Fusion power measurements provide one of the key benchmarks for any successful reactor-scale magnetic confinement fusion facility. In a deuterium–tritium fuel mix, neutrons carry about 4/5 of the produced fusion power outside the plasma volume. Thus, the 14 MeV neutron yield provides the primary direct measurement of the fusion power of a machine. The challenge of determining uncertainty in total neutron yield measurements is solved using several strategies: detailed detector assessment in a metrological neutron laboratory, in situ calibration using a mobile neutron source (typically 252 Cf), and cross-calibration using a well-characterized detector in a well-known discharge. Transition to reactor-scale devices, such as BEST, ITER, and DEMO, will require the use of neutron sources with yields of 10 10 –10 11 s −1 and above. Multiple diagnostics systems fall within this scope: neutron counters, neutron activation systems, and multi-collimator systems. Together, they provide fusion power measurements with up to 10% accuracy and 1 ms time resolution for the case of ITER, across a broad dynamic range of fusion power. Two methods for in situ calibration are considered: (1) using multiple neutron generator (NG) positions to emulate a circular or ‘ring’ source of fusion neutrons for calibrating the permanent detector set, and (2) making use of multiple temporary detectors at various locations inside the vacuum vessel during calibration, thereby allowing us to obtain more reference points for further Monte Carlo model validation with the same (or less) irradiation duration. This study details the results of neutron detector characterization under laboratory conditions using compact NGs, showing that the use of powerful (up to 10 11 s −1 D–T and 10 9 s −1 D–D) yet compact NGs with sealed tubes raises the challenge of steady-source metrological assurance, especially when considering said sources for in situ calibration.