Long-pulse Plasma Research Articles

Toroidal distribution of heat load on castellated plasma facing components for lower divertor target in EAST

In tokamak devices, the performance of plasma facing components (PFCs) under high heat loads is a key challenge for achieving long-pulse and high-performance plasma operations. The lower divertor of EAST adopts an ITER-like cassette modules (CMs) structure. However, regular distribution of damage along the toroidal direction have been observed on the divertor target plates, which not only affects the lifespan of PFCs but also limits the long-pulse and high-performance plasma operations. Therefore, investigating the toroidal distribution of the heat load on the divertor target in EAST is crucial for understanding the mechanism of damage on the target plate surface. The toroidal characteristics of heat loads on CMs of the lower divertor in EAST was analyzed using a three-dimensional magnetic field line tracing program (PFCFlux). For the inner vertical target (IVT) and outer vertical target (OVT) within a single CM, the heat load varies monotonically along the toroidal direction. In contrast, the heat load distribution on the outer horizontal target (OHT) is uniform, except at the chamfers, which is directly correlated with the target’s geometric structure and the magnetic field configuration. The surface heat load distribution exhibits toroidal non-uniformity, with the heat deposition pattern being generally consistent and displaying periodic variations. The toroidally uneven heat load distribution is attributed to the change in the magnetic field’s tilt angle. Additionally, the peak heat load is observed at the chamfer positions between cassettes, and an analysis is conducted on the relationship between peak heat load and the size of the chamfer. Smaller tilt angles result in lower surface heat load. Furthermore, a comparative analysis of peak heat loads at the chamfer under different plasma current conditions reveals that higher plasma currents generally result in increased peak heat loads. This investigation into the toroidal heat load distribution on the divertor surface contributes to understanding of the damage mechanism of W PFC as the target plate and provides valuable data for divertor design of future fusion devices.

Overview of recent experimental results on the EAST Tokamak

Since the last IAEA-FEC in 2021, significant progress on the development of long pulse steady state scenario and its related key physics and technologies have been achieved, including the reproducible 403 s long-pulse steady-state H-mode plasma with pure radio frequency (RF) power heating. A thousand-second time scale (∼1056 s) fully non-inductive plasma with high injected energy up to 1.73 GJ has also been achieved. The EAST operational regime of high β P has been significantly extended (H 98y2 > 1.3, β P ∼ 4.0, β N ∼ 2.4 and n e/n GW ∼ 1.0) using RF and neutral beam injection (NBI). The full edge localized mode suppression using the n = 4 resonant magnetic perturbations has been achieved in ITER-like standard type-I ELMy H-mode plasmas with q 95 ≈ 3.1 on EAST, extrapolating favorably to the ITER baseline scenario. The sustained large ELM control and stable partial detachment have been achieved with Ne seeding. The underlying physics of plasma-beta effect for error field penetration, where toroidal effect dominates, is disclosed by comparing the results in cylindrical theory and MARS-Q simulation in EAST. Breakdown and plasma initiation at low toroidal electric fields (<0.3 V m−1) with EC pre-ionization is developed. A beneficial role on the lower hybrid wave injection to control the tungsten concentration in the NBI discharge is observed for the first time in EAST suggesting a potential way toward steady-state H-mode NBI operation.

Multi-delay coherence imaging spectroscopy optimized for ion temperature measurements in the divertor plasma of the Wendelstein 7-X stellarator

A new coherence imaging spectroscopy (CIS) diagnostic optimized to measure the C2+ impurity ion temperature Ti spatial distribution in the divertor plasma of the W7-X stellarator is designed, tested, and validated. Using CIS to obtain Ti in the edge of magnetically confined plasmas has historically been challenging because Doppler broadening and Zeeman splitting have comparable effects on the shape of spectral emission lines. To distinguish between these two mechanisms, a novel approach to birefringent crystal design is employed to minimize the diagnostic’s sensitivity to Zeeman splitting. The recently developed pixelated multi-delay CIS approach is also used to obtain four times as much spectral information as traditional CIS approaches. The Ti-optimized CIS diagnostic is validated in a long-pulse W7-X plasma by comparison with a high-resolution spectrometer whose sightlines overlap with the CIS field of view. The CIS and spectrometer Ti profiles have the same shape and agree to within 10% on average and 25% in the worst case. Images of the Ti distribution near the divertor show toroidally elongated bands aligned with the magnetic field, with Ti ranging between 10 and 40 eV.

First realization of LHW–plasma coupling feedback control for long-pulse operation in EAST

To sustain good lower hybrid wave (LHW)–plasma coupling for long-pulse plasma operation, for the first time, coupling feedback control is designed and realized in EAST using a proportion integration differentiation method by choosing the reflection coefficient (RC) of LHW power as the reference for gas-puffing feedback, and including one pulse test and multi-pulse experiments. Experiments show that such feedback control can work correctly and maintains good LHW–plasma coupling effectively for a long time, suggesting the possibility of feedback control application on LHW–plasma coupling in long-pulse plasma. Furthermore, during the feedback control process of multi-pulse supersonic molecular beam injection (SMBI), the stored energy changes from 29 kJ to 58 kJ, and the energy confinement factor (H 89) increases from 0.98 to 1.45, implying a positive effect of coupling feedback on plasma performance. Experiments between SMBI puffing and the gas puffing system, fed by a piezoelectric valve near the antenna, are further investigated, showing that the response time of the RC with SMBI is faster than that by the piezoelectric valve. In addition, SMBI puffing on the electron-drift side of the LHW antenna is a little quicker than that on the ion-drift side. Studies suggest that such feedback control is effective for long-pulse LHW–plasma coupling, and the gas puffing by SMBI on the electron-drift side of the LHW antenna could offer an effective way to sustain good LHW coupling in steady-state operation in the future. Further optimization will be continued at a later date.

Progress of CO2 Dispersion Interferometer on EAST

A CO2 Dispersion Interferometer (DI) system on the Experimental Advanced Superconducting Tokamak (EAST) was successfully operated, providing plasma electron density measurements. The DI system utilizes a continuous-wave 9.3 μm CO2 laser source to measure line-averaged electron densities. This offers significant advantages, including the ability to minimize fringe jumps and insensitivity to mechanical vibrations. These characteristics are well-suited for future high-density, long-pulse plasma discharges. The DI system provides a real-time density feedback signal to the plasma control system for routine density control during long-pulse operation. Experiments with EAST demonstrated good agreement between the density obtained by the DI system and the preset densities. The DI system also exhibited stability during long-pulse discharge. Moreover, the DI system was stable during rapid density changes and high-density pellet injections. In shot No. 120594, the DI system exhibited stable density feedback during continuous projectile injection lasting over 50 seconds; the line-averaged electron density is approximately 4×1019 m-3. In contrast to the long-wavelength source interferometer, which may deflect light from the detector owing to excessive refraction angles in larger density-gradient discharges, the DI ensured accurate density measurements. The DI system on EAST is dependable for accurately measuring the electron density.

Design and analysis of a PAM launcher at 4.6 GHz for a new LHCD system on EAST

To improve the Current Drive (CD) capability in long-pulse (up to ∼1000 s) H-mode operation, it has been decided to develop a new Lower Hybrid Current Drive system at 4.6 GHz with an active cooling Passive Active Multijunction (PAM) launcher on EAST. In this paper, both the radio frequency (RF) and the physical properties of this PAM are studied numerically. The same nominal parallel refractive index (N || = k ||c/ω, where k || is the parallel wavenumber, c the velocity of light, and ω the wave angular frequency) of 2.04 as the existing 4.6 GHz Full Active Multijunction (FAM) is chosen. Ray-tracing calculations indicate that good accessibility could be achieved when the LH waves radiate with this nominal N || in typical long-pulse H-mode plasmas. The coupling performance in terms of power reflection coefficient (R C), power spectrum, maximum electric field, power directivity (D P) and global CD capability is evaluated with the ALOHA code based on the linear coupling theory. Good coupling performance with averaged R C ⩽ 1% and D P ∼ 70% could be expected with the density (n e) in front of the PAM close to the cut-off value (n e_co). The simulated R C remains below 6.5% over a wide density range 0.5 ⩽ n e/n e_co ⩽ 10, which is similar to the plasma edge conditions produced by Edge Localized Mode activity. A detailed comparison with the existing 4.6 GHz FAM launcher is also performed.

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.

Design and development of the EAST plug-in cryopump for long-pulse high-performance operation

At present, the in-vessel cryopump of the EAST divertor demonstrates inadequate pumping performance during long-pulse high confinement mode (H-mode) plasma operations, and the disassembly and maintenance of this cryopump pose significant challenges. In response to these issues and considering the constraints of the installation space of the cryopump, an EAST plug-in cryopump have been developed based on cryoadsorption and cryocondensation principles. The plan is to install four plug-in cryopumps at different vertical ports of the EAST vacuum vessel. These cryopumps will operate in a staggered manner with a specific time interval to achieve a quasi-continuous pumping system. This paper provides a systematic discussion of the design, simulation analysis, and experimental studies of the plug-in cryopump to validate its feasibility for the EAST application. In conclusion, each individual cryopump exhibits an average pumping speed of approximately 9255 L/s for hydrogen, 7001 L/s for deuterium, and 5290 L/s for helium. The pumping capacity and regeneration time also meet the requirements of EAST experiments. Our ultimate goal is to achieve the long-pulse high-performance plasma operation in EAST with the support of the plug-in cryopumps, while also laying the foundation for the development of divertor cryopumps for future fusion reactors.

Energy efficiency of N2 vibrational excitation in the spark-and-glow phases of a long-pulse air discharge at atmospheric pressure

The vibrational excitation of N2 molecules has a significant impact on plasma chemical synthesis, including nitrogen fixation and ammonia formation, by reducing the energy barrier of chemical reactions. However, experimental data for energy efficiency of N2 vibrational excitation is very rare. In this work, vibrational coherent anti-Stokes Raman scattering was applied to a 200 μs long-pulse air plasma, and the time behaviours of rotational and vibrational temperatures of N2 in two different discharge modes, namely the spark and glow modes in the long discharge pulse, were quantified. It reveals that the spark discharge produced a vibrational temperature exceeding 5000 K but with a gas temperature as low as 400 K, while the glow discharge resulted in efficient gas heating with a gas temperature as high as 3500 K and a vibrational temperature higher than 5000 K. Combined with the 0D simulation of the air plasma, it demonstrates that the energy efficiency of N2 vibrational excitation in the glow phase is about three times higher than that of the spark phase. These results provide essential inspiration for further studies of plasma chemical synthesis dominated by N2 vibrational excitation.

On the frequency bifurcations of the MHD startup modes in NSTX

The observed bifurcations of the low frequency (&amp;lt;50 kHz) and low toroidal periodicity (n &amp;lt; 5) magnetohydrodynamic (MHD) activity often present in the initial part of the National Spherical Tokamak Experiment (NSTX) discharges can be explained by the evolution of the radial profile of the safety factor (q=rBϕ/RBθ) crossing multiple rational surfaces in the core. Important performance limiting instability mechanisms in the NSTX spherical tokamak are often linked to low frequency and low-n MHD activity. They are quite common in long-pulse NSTX plasmas. They can be present at the beginning of the plasma current flat-top, at the end of the discharge or during the whole duration, and they have been observed to deleteriously impact performance over a wide range of q95. An interesting feature observed in some NSTX discharges is the presence of a bifurcation in the frequency of the low n modes, as low as n = 1, that have frequencies comparable to the plasma core rotation divided by n. Equilibrium reconstructions constrained by magnetic diagnostics data and motional stark effect pitch angle radial profiles suggest that the observed bifurcations are linked to a fast evolving minimum value of q. 3D non-linear resistive MHD simulations show that these modes are ideal and exist as non-resonant before the correspondent rational surface enters the plasma.

Long-pulse high-performance H-mode plasmas achieved on EAST

A record duration of a 310 s H-mode plasma (H98y2 ∼ 1.3, ne/nGW ∼ 0.7, fBS &amp;gt; 50%) has been recently achieved on experimental advanced superconducting tokamak (EAST) with metal walls, exploiting the device's improved long-pulse capabilities. The experiment demonstrates good control of tungsten concentration, core/edge MHD stability, and particle and heat exhaust with an ITER-like tungsten divertor and zero injected torque, establishing a milestone on the path to steady-state long-pulse high-performance scenarios in support of ITER and CFETR. Important synergistic effects are leveraged toward this result, which relies purely on radio frequency (RF) powers for heating and current drive (H&amp;CD). On-axis electron cyclotron heating enhances the H&amp;CD efficiency from lower hybrid wave injection, increasing confinement quality and enabling fully non-inductive operation at high density (ne/nGW ∼ 70%) and high poloidal beta (βP ∼ 2.5). A small-amplitude grassy edge localized mode regime facilitates the RF power coupling to the H-mode edge and reduces divertor sputtering/erosion. The high energy confinement quality (H98y2 ∼ 1.3) is achieved with the experimental and simulated results pointing to the strong effect of Shafranov shift on turbulence. Transport analysis suggests that trapped electron modes dominate in the core region during the record discharge. The detailed physics processes (RF synergy, core-edge integration, confinement properties, etc.) of the steady-state operation will be illustrated in the content. In the future, EAST will aim at accessing more relevant dimensionless parameters to develop long-pulse high-performance plasma toward ITER and CFETR steady-state advanced operation.

Optical plasma boundary detection and its reconstruction on EAST tokamak

Plasma boundary detection and reconstruction are important not only for plasma operation but also for plasma facing materials. Traditional methods, for example, EFIT code, which is constrained by electromagnetic measurement, and is very challenging for detecting the plasma boundary in long-pulse burning plasma devices such as ITER. A novel algorithm for the reconstruction of the plasma boundary using one visible camera has been developed on experimental advanced superconducting tokamak (EAST) for fusion reactors. A U-Net convolutional neural network was used to identify the plasma boundary and the pixel coordinates of the boundary points were fitted with EFIT via the XGBoost model. This algorithm can transform the boundary from the image plane to the poloidal plane of the Tokamak based on machine learning without traditional spatial calibration, and then the reconstruction of the plasma configuration shall be realized based on a monocular visible light camera. The reconstruction accuracy of this algorithm is relatively high. The average error on the test set was only 7.36 mm (<1 cm) and satisfied the accuracy requirements of control for EAST tokamak. This result can contribute to the development of the plasma boundary reconstruction and operation based on one visible camera.

Heat transfer analysis of the baking system for EAST

Before long-pulse plasma operation, wall conditioning is addressed as a key element to achieve steady state long-pulse operation with high performance, which is one of the major challenges for EAST (Experimental Advanced Superconducting Tokamak). The vacuum vessel and in-vessel components are continuously heated with hot nitrogen for a few weeks, namely the baking process, and the beginning of wall conditioning. It is difficult to accurately predict the specific temperature distribution of the baking system in the baking process under the current analysis and experimental conditions. Therefore, a simplified calculation model has been established in this paper, which is helpful to reveal the dynamic response mechanism and the heat loss to guide the input power of the baking system. The heat transfer relationship has been given by the theoretical method, and the calculation results of two cases are obtained by using Matlab/Simulink. The results gained by the heat transfer model were compared with the actual experimental data, which verified the correctness of the model. This heat transfer calculation model can be used in the future EAST experiment to predict the baking system operation behavior.

The first achievement of the double feedback control of the detachment in the long-pulse plasma on EAST

The first achievement of the double feedback control of the detachment in the long-pulse plasma on EAST

In situ melting phenomena on W plasma-facing components for lower divertor during long-pulse plasma operations in EAST

Tungsten (W) is one of the most promising plasma-facing materials for future fusion devices. Although its melting point is the highest among all metals, it still has great risk of melting under extremely high plasma heat fluxes, which is a big concern for ITER and future reactors. Actively cooled W plasma-facing components (PFCs) with both monoblocks and flat-type structure have been successfully installed in the lower divertor of the EAST tokamak since 2021, and provide a good opportunity for direct comparison of the damage mechanism for the two types of PFCs. Various in situ melting phenomena on the lower divertor have been observed by CCD cameras, which have been further verified by post-mortem inspections. Severe melting and even exfoliation of the edge-beveled W plates were observed on some W/Cu flat-type components at horizontal outer targets. Many droplets were ejected during long-pulse operations, which induced a significant increase of W impurities and total irradiation in the core plasma, and thus greatly deteriorated the plasma performance and even caused disruptions. Two different shaping structures of flat-type PFCs show different positions of melting and the corresponding mechanisms. Slight melting was found on the sharp leading edges of W/Cu monoblocks between cassette modules (inter-CM) for horizontal targets with small droplet ejection, which was much improved compared to that observed on the upper W divertor, illustrating that the application of a large-sized bevel chamfer inter-CM was generally effective. In addition, an unexpected melting phenomenon on the dome plate was attributed to the extreme transient heat flux during disruption with runaway electrons. The application of both types of W/Cu PFCs for the divertor provides important experiences and lessons for the engineering design and optimization of divertor PFCs in future fusion devices.

Realization of thousand-second improved confinement plasma with Super I-mode in Tokamak EAST

Mastering nuclear fusion, which is an abundant, safe, and environmentally competitive energy, is a great challenge for humanity. Tokamak represents one of the most promising paths toward controlled fusion. Obtaining a high-performance, steady-state, and long-pulse plasma regime remains a critical issue. Recently, a big breakthrough in steady-state operation was made on the Experimental Advanced Superconducting Tokamak (EAST). A steady-state plasma with a world-record pulse length of 1056 s was obtained, where the density and the divertor peak heat flux were well controlled, with no core impurity accumulation, and a new high-confinement and self-organizing regime (Super I-mode=I-mode+e-ITB) was discovered and demonstrated. These achievements contribute to the integration of fusion plasma technology and physics, which is essential to operate next-step devices.

Recent EAST Experimental Results and Systems Upgrade in Support of Long-Pulse Steady-State Plasma Operation

In recent years, EAST has achieved a 60-s scale steady-state plasma operation (SSO), which is electron dominant heated by the electron cyclotron wave and low hybrid wave. Substantial progresses have been made in the development of long-pulse operation, such as active control of radiative divertor, edge localized mode (ELM) suppression, and He plasma experiment. To further develop the steady-state operation scenario and study the related critical physical issues, the upgrade of EAST has been completed in 2020. The auxiliary heating systems were rearranged and upgraded with a total heating power capability of 33 MW. The 2.45-GHz lower hybrid current drive (LHCD) launcher has been updated from full active multijunction (FAM) to passive active multijunction (PAM) to avoid damage caused by plasma–material interaction. The injection power capability of electron cyclotron resonance heating (ECRH) system is upgraded to 1.4 MW. The upgraded ion cyclotron resonance frequency (ICRF) antenna injects low parallel wavenumber ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$k_{\vert \vert }$ </tex-math></inline-formula> ) spectra with good single pass absorption. The neutral beam injection (NBI) system is moved from F-port to D-port and is reoriented from counterclockwise to clockwise injection. The armor material of lower divertor has been upgraded from graphite to tungsten improving the steady-state heat handling capability from 2 to 10 MW/m2. Based on the upgraded pumping system and improved conductance, the effective pumping speed for <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$D_{2}$ </tex-math></inline-formula> of the EAST lower divertor increased twice. With all these upgrades, 100-s scale steady-state operation ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$T_{e0} &gt;10$ </tex-math></inline-formula> keV) and 1056-s high-performance operation with 1.73-GJ injection energy were achieved in 2021 campaigns. The experience in high-performance long-pulse operation obtained on EAST will lay a foundation for the ITER and Chinese Fusion Engineering Testing Reactor (CFETR) operation.

Non-inductive plasma vertical position measurement for the 1056s discharge on EAST.

Vertical position stability plays a crucial role in maintaining safe and reliable plasma operation for long-pulse fusion devices. In general, the vertical position is measured by using inductive magnetic coils installed inside the vacuum vessel; however, the integration drift effects are inherent for steady-state or long-pulse plasma operation. Developing a non-magnetic approach provides a fusion reactor-relevant steady-state solution that avoids the negative impact of integration drift. In this paper, we compare the non-inductively determined vertical position achieved by line-integrated interferometer and polarimeter measurements to that employing an inductive flux loop for a 1056s discharge recently achieved on EAST (Experimental Advanced Superconducting Tokamak). Experimental results show that the non-inductive measurement is more robust than flux loops after 300s if the integrator is not reset to suppress integrator drift. Real-time vertical position control using the non-inductive system is proposed for the next EAST experimental campaign.

Equilibrium effects on the structure of island divertor and its impact on the divertor heat flux distribution in Wendelstein 7-X

To support the planning of high-performance long-pulse plasma operation for W7-X, the role of equilibrium effects for the structure of island divertor and its profound impacts on the divertor heat flux distribution have been investigated. Finite-beta full-field three-dimensional equilibria for the four most important configurations of W7-X (‘standard’, ‘high-mirror’, ‘high-iota’ and ‘low-iota’) were obtained with the HINT code. Equilibrium calculations show a strong dependence of the edge stochastization on the magnetic configurations. Parameterization to estimate equilibrium effects on edge stochastization and plasma shape was attempted for individual functions. The results indicate that a more stochastic boundary can be attained with increasing plasma beta, except for the low-iota configuration. According to fully diffusive heat-load estimations, it turns out that equilibrium effects are beneficial for spreading the heat flux on the divertor target. The interactions of the beta effects with effects driven by the toroidal plasma current are also analyzed for the standard configuration. Infrared camera measurements confirm the predictions concerning global heat load distributions.

First lithium experiments in HIDRA and evidence of helium retention during quasi-steady-state stellarator plasma operations

Recent experiments in Hybrid Illinois Device for Research and Applications (HIDRA) have had operational discharges between t discharge = 60 and 1000 s using electron cyclotron resonant heating (ECRH) of the plasma. This means that quasi-steady-state plasma discharges reach conditions to study long-pulse plasma material interactions (PMIs). The newly commissioned HIDRA-Material Analysis Test-stand PMI diagnostic is used to place a drop of lithium onto a heated tungsten surface, transfer the sample in-vacuo and expose it in a helium plasma. Helium is of interest as there is an open question to whether lithium will be able to remove helium ash in real fusion devices. The introduction of the W-Li sample in HIDRA resulted in evaporation of lithium into the helium plasma during a 600 s pulse and caused a reduction of over 90% in neutral pressure during the discharge. It was also observed that the plasma density and temperature increased by over 2.5 times. Using spectroscopy and a helium collisional radiative model, the peak temperature and density of the helium plasma can be monitored during the discharge. During lithium evaporation, as significant lithium ionization occurs, there is a 85% drop in the HIDRA vessel neutral pressure, despite a constant flow rate of He gas. This reduction in neutral pressure is supported by spectroscopy data with corresponding reductions in He I line intensities (587 nm, 667 nm, 706 nm, and 728 nm), as well as those of other impurities. At one point in the discharge a lithium plasma is created, as indicated by an increase in Li+ emission and a complete reduction in He+ emission, but the electron density jumps from n e = 3 × 1018 m−3 to over n e = 8 × 1018 m−3 while the core temperature stays relatively constant between T e = 16 eV and 20 eV. Once lithium has completely evaporated from the sample and the majority of the ionized lithium has diffused from the plasma to the vessel walls, pressure and spectroscopy data paired with He collisional radiative model calculations shows a re-establishment of a helium plasma in a low recycling regime. In this regime, the density drops down to n e = 2 × 1018 m−3 and the electron temperature increases from T e = 20 eV to over T e = 50 eV indicating an increase in helium heating efficiency. This is also indicated by the He+ emission re-establishing and having a higher intensity. In this paper we show the results from the first lithium campaign in HIDRA. In the presence of lithium, and in particular when lithium ions are present, the helium disappears from the plasma via an as of yet unknown complex relationship that needs to be further studied. The most likely explanation is that the lithium ions are distributed around the vessel and able to trap helium to the surface turning HIDRA into a large gettering surface. These results have potential implications on future plasma facing component design using liquid lithium for impurity and recycling control using limiters and divertors.