Runaway electron mitigation with pulsed localized vertical magnetic field perturbation in ADITYA tokamak

To reduce the risk of severe damage to the vessel and inner peripherals of any tokamak and its safe operation, a robust technique for the mitigation of runaway electrons (REs) is required. The REs in ADITYA tokamak are effectively mitigated by an application of local vertical magnetic field (LVF) perturbation. The LVF perturbation is applied using a pair of electromagnetic coils placed at the top and bottom of the ADITYA vacuum vessel in a Helmholtz configuration at one toroidal location. Powered by a capacitor bank power supply, these coils can produce a localized vertical magnetic field at the plasma center in the range of ∼150 G–260 G for a variable duration of 5–20 ms. The LVF pulse is first applied at the breakdown/current-ramp phase, where the REs are generated in the discharges initiated by the conventional ohmic breakdown in ADITYA. With the application of LVF pulse the REs are significantly reduced as indicated by the reduction in the REs generated hard x-ray flux. It has been observed that to extract the REs efficiently, an LVF pulse of magnitude at least ∼1% of the toroidal magnetic field with a minimum duration of ∼5 ms should be applied. The LVF perturbation is applied at different times into the discharge, i.e. during the breakdown/current ramp-up phase and current flat-top phase. The REs are significantly reduced in all the phases and improved discharge consistency. The LVF acts as an error field and a short-pulse of the LVF influences the REs more in comparison to the thermal electrons due to the faster velocities of the REs.


Introduction
Runaway electrons (REs) in tokamaks are highly energetic electrons that run away in the velocity space due to driving parallel electric field force, which conquers the friction force due to collisions [1]. These electrons are accelerated by the toroidal electric field generated by ohmic transformer coils as well as the electric field generated during the current quench phase of disruption due to rapid plasma current decay rate (dI P / dt) [2]. REs are observed in early experiments in low-density discharges contaminated with impurities and have later been studied experimentally in more detail in several devices. The control of REs in tokamak plasma discharges remains a serious concern for reliable tokamak operation, because a significant fraction of REs may restrict the efficiency of ohmic heating and prevent the creation of a hot fusion plasma. Most of the complex plasma physics processes such as MHD instabilities, plasma disruption, particle transport, plasma equilibrium control, impurity recycling, and field perturbation effects are influenced by REs [3]. REs avoidance/suppression remains one of the exceptional problems for future burning plasma experiments including ITER [4] for their safe and reliable operation. ITER start-up scenario includes plasma formation that will be achieved with the low prefill pressure (because of the lower availability of electric field (0.3 V m −1 ) for plasma breakdown), remains prone to becoming REs-dominated discharges due to poor collisionality [5]. High energetic electrons with energies of tens or hundreds of MeV in large-size tokamaks such as JET [6], and ITER [7], can carry a significant fraction of plasma energy. This poses a severe threat to the peripheral plasma-facing components as well as to the vacuum vessel of a tokamak machine [8] if locally deposited during plasma current disruption. The REs generated during a disruption is also a threat to the machine's safety. The thermal quench significantly increases the resistivity on a time scale much shorter than the magnetic diffusion time scale, thereby increasing the ohmic electric field to drive REs. The following current quench generates a large inductive electric field that drives additional REs, in addition to the avalanche source of secondary REs [2]. This is the main reason why the REs suppression /mitigation is an essential part of tokamak research particularly for ITER and in future commercial nuclear fusion reactors [9].
There are several techniques developed and tested in various devices around the world for high-energy RE mitigation. The RE mitigation during disruption under the influence of injected impurities such as neon, argon through massive gas injection (MGI) has been demonstrated on TEXTOR [10], ASDEX-U [11,12], DIII-D [13], and JET [14]. Experiments on generation of RE beam triggered by a massive injection of noble gases like argon or neon using two different scenarios such as the ramp-up scenario and the flat-top scenario have been studied in the COMPASS tokamak [15]. However, certain limitations of the MGI technique, such as the slow flow response time, and poor gas penetration of the impurities injected have been overcome through shattered pellet injection (SPI). The SPI enables ITER-relevant injection technique, in which high Z (neon, argon, etc) cryogenic impurity pellets are injected. The RE current has been successfully dissipated using SPI in DIII-D [16] and J-TEXT [17]. Another important method that was implemented for REs reduction is based on resonant magnetic perturbation (RMP). The REs reduction using RMP in the flat-top and the disruption were demonstrated on TEXTOR [18,19] and JT-60U [20] respectively. The REs mitigation using RMPs with dominant toroidal mode number n = 1 generated by external coils applied during the pre-disruption phase and post-disruption phase have been studied experimentally in ASDEX UPGRADE [21]. The 3D field-induced RE loss has been simulated for DIII-D and COMPASS plasmas, utilizing the MARS-F code incorporated with the updated REORBIT module [22]. However, Riccardo et al [23] and Papp et al [24] have reported that REs mitigation using RMPs does not seem to be an effective technique for large-size devices such as JET or ITER. This is due to the fact that the required short-scale perturbations decay rapidly with the distance from the coils and because of the large minor radius of these devices, the losses of REs are limited from the region close to the edge of the plasma. The perturbations might be insufficient to expel the particles from the core. Furthermore, the REs suppression during the flattop on account of a relative decrement in the loop voltage was attained using ECRH and LHCD in FTU [25] and for LHCD in HT-7 [26]. During early experiments on REs reduction, the magnetic extraction of REs from a tokamak plasma has been reported on a small research tokamak, VERSATOR-I [27]. ADITYA/ADITYA-U being medium-size tokamak with moderate plasma currents (90-200 kA) and plasma duration (100-350 ms), REs do not attain catastrophic energy levels and hence they can be studied in a controlled manner. Previously, the primary (Dreicer mechanism) REs suppression using an additional gas puff of working gas in TCABR [28] and ADITYA [29,30] /ADITYA-U [31] and later using SMBI [32] has been reported and numerical studies for REs suppression in ADITYA-U has been performed [33]. In this work, we report a detailed investigation of an efficient technique for magnetic extraction of REs that was earlier applied to the VERSATOR-I tokamak [27] having relatively low-level plasma parameters and small machine size. In this technique, a local vertical magnetic field (LVF) has been applied in the direction opposite to the main vertical magnetic field in ADITYA tokamak [34] during the early phase of discharge time (0.5-15 ms) that leads to an immediate escape/extraction of REs before they gain high energies and decrease the efficiency of ohmic heating. In addition, we find that this RE extraction technique is efficient for the plasma discharge having a higher impurity recycling from the first wall surface, while the gas-puff technique does not provide better control on the REs because of poor heating of the plasma. In this case, the LVF perturbation technique has been found to be the most suitable technique for the extraction of REs even during all phases, i.e. start-up, flat-top, and disruption phases, of the discharge. The experimental set-up along with LVF coils' dimension details are discussed in section 2. The magnetic field mapping for LVF coils and its comparison with simulation are reported in section 3. The experimental observations and results of REs suppression on ADITYA tokamak are discussed in section 4. Section 5 summarizes the article with a future plan.
The major diagnostic systems used in these experiments include magnetic sensors such as external loops for loop voltage (V loop ), software as well as hardware integrated Rogowski coils for plasma current (I p ), and in-vessel position probes for plasma position (∆R, ∆Y) measurements. Visible spectrometers and photomultiplier tubes (PMTs) with wavelength filters are used to measure line radiations such as neutral (H α ) and O-I, C-III, visible continuum impurities. The hydrogen fuel pressure is monitored and controlled using a piezo-electric valve operated in a pulsed (pre-filled) gas-feeding configuration. The chord averaged electron density and the central electron temperature are measured using microwave interferometry and soft x-ray detectors respectively. The presence of REs in a tokamak is normally identified by the intensive bursts of hard x-rays (HXRs) when these fast electrons collide with a thick target (such as a limiter or chamber wall) [35]. The HXRs counts are monitored in this experiment using NaI (Tl) scintillator detectors of two different diameters i.e. 1.5 inches and 3 inches, operated in a current mode with PMT readout. The lead (Pb) shielded scintillator is positioned at the machine mid-plane (Z = 0) and collimated to view the HXR radiation from a cone of plasma and limiter. The thick target bremsstrahlung radiation coming from the limiter is assumed to be dominant [36][37][38].
A pair of Helmholtz-like REs extraction coils are designed, manufactured, and symmetrically placed at one toroidal location on the top and bottom of the TF coil I-Beam (near limiter location above port# 2 & 3 of vacuum vessel) of ADITYA. A schematic diagram of a pair of REs extraction coils along with their dimension details is shown in figure 1(a). The photographs of both the REs extraction coils are shown in figures 1(b) and (c).
The distance from the major axis to the coil center (R) is ≈ 870 mm and the vertical distance from the mid-plane (Z = 0) to the coil center (Z) is ≈ ±820 mm. The inner diameter and the outer diameter of the coils are ≈424 mm and ≈510 mm respectively. The radial width (∆r) is ≈ 43 mm, the vertical width (∆Z) is ≈ 85 mm, the conductor diameter (⊘) is ≈ 8.5 mm, and the number of turns (n) is ≈ 50 turns/coil. The plasma current (I P ) direction in ADITYA is clockwise as seen from the top. The actual vertical magnetic field (equilibrium field) is directed from bottom to top (upward) as seen from the top. The LVF perturbation field is directed top to bottom (antiparallel) to the actual equilibrium field. The coils are powered by a capacitor bank power supply (fast bank 1 mF/10 kV, slow bank: 14 mF/5 kV with crowbar circuit). The current is monitored with C.T (10 mV/1 amp). The current pulse length is 5 ms without the crowbar and it is ∼20 ms when the crowbar is operational.

LVF measurements
The requirement of the magnitude of the LVF perturbation field at the plasma center is estimated from the previous experiments reported in [27] for efficient REs extraction. Taking the ADITYA tokamak operating parameters (B T ∼ 0.75 T to 1.5 T, I P ∼ 100-200 kA), the requirement of LVF perturbations field is estimated to be ≈150-300 G at the plasma center for efficient REs mitigation. For obtaining the above-mentioned perturbation field at the plasma center, the coil dimensions are chosen based on the ADITYA machine geometry and the space availability between two toroidally separated ports located at the top and bottom of the machine. For placing the coils as close as to the vacuum vessel, this space is required. The optimal configuration is arrived at by experimenting with two different sets of LVF coils.
For the ADITYA tokamak, the LVF perturbation field produced by a pair of REs extraction coils (as shown in figure 1) is measured using a Digital Gauss Meter as a function of distance in the vertical (Z) direction from the outer periphery of the vessel (Z = +30 cm) to the coil center. At the outer periphery of the vessel, the magnitude of the perturbation field is ∼100 G at ∼1.7 kA of peak coil current, which maximizes to ∼260 G at ∼4.2 kA of peak coil current. Furthermore, the estimation of the magnetic field produced by a pair of REs extraction coils in Helmholtz coils-like configuration has been modeled using COMSOL Multiphysics, commercially available software [39], to obtain the field inside the vessel up to its center.
where A-Magnetic vector potential, V-Electric scalar potential, J e -Externally applied current density vector, σ-Electric conductivity, µ 0 -Permeability in a vacuum. The above magneto static equation is solved using the finite element method in the 2D-axisymmetric model to find the magnetic field in the neighborhood with a fine mesh (mesh quality >0.9) and using infinite elements at the boundary.

Experimental observations
Under the experimental conditions of ADITYA operation, a large fraction of non-thermal (runaway) electrons are commonly observed in the early phase of the discharges with a higher toroidal electric field, low densities, and /or high level of impurities. The nature of REs-dominated ADITYA plasma discharges is similar to REs-dominated discharges observed in other tokamaks. Generally, current steps, loop voltage spikes and a corresponding spike in neutral and other impurity line radiation, inward shifts in the radial plasma position, and an intensive burst of HXRs, etc are commonly observed in these types of RE-dominated discharges. Figure 4 shows the temporal evolution of plasma parameters that represent the typical REs-dominated ADITYA discharge (#23615). The typical loop voltage spikes [36] in figure 4(a) and corresponding spikes in neutral (H α ) ( figure 4(d)) are observed. The loop voltage spikes are related to the RE current as explained using a one-dimensional (1D) diffusion model of the toroidal electric field, which is produced due to RE beam movement inside the plasma. The RE beam movement inside the plasma and its subsequent loss induced a positive electric field at the radial location from where they lost and induced a negative toroidal electric field at the locations where they arrive inside the plasma. These negative and positive electric fields then diffuse out to the plasma boundary to appear as negative and positive loop voltage spikes [36]. The plasma current steps corresponding to loop voltage spikes are visible in figure 4(b). The inward shift of the radial plasma position (figure 4(f )) occurred at 15.7 ms time when the first negative loop voltage spike appears. The inward shift in the plasma column brings more impurity from the wall that increases the impurity recycling after 20 ms of the discharge length. The rapid fall in the electron density after 20 ms is observed in figure 4(c) and keeps at density ≈4 × 10 18 m −3 from 40 ms onwards, which is below the critical density that converts the discharge into REs regime and brings lots of HXRs from 40 ms onwards is shown in figure 4(e). Normally, ADITYA discharges are operated with pulse (pre-filled) gas feed mode. Initially, density is kept low at 1 × 10 19 m −3 , later it is raised by 1.5-2 times with hydrogen gas puffs, which suppress the HXRs throughout the discharge length (except HXR presented during the first 15 ms).
However, if the higher impurity recycling from the wall or limiter surface is dominated, the gas-puff technique will not be helpful to control the REs because of the poor heating of the plasma. In this case, the LVF perturbation technique has been found to be the most suitable technique for the extraction of REs even during the early phase (start-up) of the discharge also. A detailed experiment on the effect of LVF perturbation on the signature of REs (HXRs) and other plasma parameters is performed.

Start-up phase generated REs extraction by the LVF
The REs suppression experiments in ADITYA by using the LVF perturbation technique initially attempted to mitigate the REs during the initial phase of plasma discharges (during the first 15 ms of discharge time after the breakdown). Application of a short localized vertical field perturbation, varying it from 150 G to 300 G at a fixed time and later, the LVF pulse starting-time (LVF initiation time) varied from 4 ms to 0.5 ms at a fixed magnitude of ∼260 G [34]. Furthermore, an experimental scan for ADITYA discharges has been carried out to find the critical threshold value (least magnitude) of LVF perturbation required for efficient mitigation of the REs. Hence, the reduction in HXRs flux signal is accounted as suppression of REs. The time-integrated HXRs flux signal over the first 15 ms of the discharge length is plotted as a function of peak LVF perturbation field strength in figure 5, representing the critical threshold of LVF perturbation amplitude. The vertical dotted line in figure 5 shows the critical threshold of the LVF magnitude ∼100 G, which is of the order of 10 −2 B T .
It is clear from the above figure that the LVF perturbation amplitude, applying beyond the critical threshold has the same effect on REs suppression as that of the critical threshold value. Further, an increase in the LVF perturbation up to ∼280 G, showed a significant reduction, almost 80% in initial HXRs observed in both the detectors in discharge, when the LVF perturbation is applied. Figure 6 shows the time traces of ADITYA discharges (#24969, without the LVF perturbation and #24978 with the LVF perturbation) comparison for the plasma parameters (a) loop voltage (V), (b) plasma current (kA), (c) chord average electron density (n e ), (d) HXRs (a.u.), (e) HXRs flux (a.u.) and (f) LVF perturbation field (Gauss) during the first 15 ms of the discharge length. As seen in figure 6(b), the plasma current is observed to be reduced by a large amount during the start-up phase with the application of LVF. This phenomenon has been observed in several discharges. This suggests that the REs constitute a large fraction of current in the start-up phase. With a high value of E/ED (ratio of an applied electric field to the Dreicer electric field) and a low value of electron density (n e ) substantial fraction of RE generation occurs in the case of a conventional tokamak start-up. The modeling of the ADITYA tokamak startup phase, including realistic 1D effects within the broad 0D framework, where, the effect of REs is also incorporated, has reported RE current (I RE ) contribution of ⩾30 kA in the total current during the plasma start-up phase [40].
Later, starting time of the LVF perturbation has been varied, which showed that the perturbation reduces HXRs during any time of the discharge length. During the activation of the LVF pulse, there is no evidence of spikes in the loop voltage nor is a plasma current disruption event observed. Therefore, the LVF perturbation applied of the order of magnitude ∼3%-4% of the operating toroidal magnetic field (B T ) can reduce the HXRs or extracts the REs without affecting the thermal component of the plasma.
The REs extraction mechanism by employing the LVF perturbation is quite simple. Due to the LVF perturbation, the particle moves in a vertical direction in the neighborhood of the perturbation field. This leads to a radial diffusion, which will be proportional to the particle velocity parallel to the total magnetic field. The estimation of radial diffusion due to the LVF perturbation field is given by the equation [27], Here, V ∥ = particle velocity along magnetic field ≈ 10 8 m s −1 ; B LVF = local vertical field perturbation ≈260 G; B = Toroidal magnetic field = 7500 G. We observe, the ratio BLVF B ≈ 10 −2 is the critical threshold required to successfully mitigate the RE produced during the start-up phase of plasma discharge. With L = scale length of the perturbation field gradient ≈ 0.1 − 0.2 m, D ⊥ ≈ 250-1000 m 2 s −1 . As the REs have higher parallel velocity by an order of magnitude as compared to thermal particles, the REs diffusion will be larger (∼10 times) than the thermal particles. Thus, the REs suppression can occur without affecting the thermal component of the plasma. Note here that the Rechester-Rosenbluth model [41] for RE diffusion is mainly associated with particle and heat diffusion in stochastic magnetic fields. Furthermore, Särkimäki et al [42] have reported an advection-diffusion model for RE diffusion which is again based on field stochastization and is slightly different from Rechester-Rosenbluth diffusion formula. As the application of the LVF perturbation field may not be leading to field stochastization [43], it may not be appropriate to use the Rechester-Rosenbluth formula in the present case.
An estimate of the minimum required δBLVF BT for RE mitigation can be made, by equating the complete RE loss rate by the applied LVF perturbation to the RE generation rate due to primary and secondary RE generation mechanisms Here, τ δB is the characteristic diffusion time associated with magnetic turbulence, and τ dr −1 and τ av −1 are the RE generation time scales of primary and secondary mechanisms respectively.
where v ∥ is the parallel electron velocity, γ (≈ 3) is the relativistic scaling factor and D M is the magnetic diffusion coefficient [44]. The 0D model of the RE generation and magnetic perturbation-induced RE loss can be expressed as [44,45] ∂n RE ∂t = Here τ dr is the RE generation rate due to the primary or Dreicer mechanism, τ av is the RE generation rate due to the secondary or avalanche mechanism and τ loss is the RE loss rate due to magnetic field perturbation. This generation and loss rate of REs determines the temporal evolution of RE densities. The REs generation rate due to the primary mechanism [46] is expressed as, And, the REs generation due to a secondary mechanism (i.e. avalanche mechanism) can be expressed as below, where ϵ d is the ratio of E/E C , where E is the toroidal electric field and E c is the critical electric field. The critical electric field E c is expressed as E c = e 3 nelnΛ 4 π ϵ 2 0 mev 2 th . Where ν e is the collision frequency of the electrons at the thermal velocity v th , e is the charge of an electron, m e is the mass of an electron, ϵ 0 is the permittivity of free space, lnΛ is the Coulomb logarithm, and Z eff is the effective plasma ionic charge [44]. The constant factor k in equation (5) is taken to be ∼0.43 [3].
The values of the parameters used in the calculation are shown in the following table 1.   We find the calculated critical threshold δBLVF BT ⩾ 10 −2 , for the 100 keV energetic electrons generated during the startup phase is in good agreement with the experimental critical threshold.

Discharge parameters improvement
Detailed experiments revealed that the LVF pulse duration (width) has to be relatively increased as compared to the required pulse duration for REs extraction during the start-up phase, for improving the discharge behavior. Hence, the LVF current pulse length has been prolonged by adding another capacitor bank in operation named a slow bank (10.5 mF/5 kV) capacitor power supply, which has been simultaneously fired with the fast bank. Figure 7 illustrates a comparison of the LVF current pulses with and without the slow bank capacitor power source. The load is crowbarred at the peak of the current in both current pulses so that the capacitors do not experience any undesirable voltage reversal on the capacitors. The modified current pulse shown in figure 7 with magenta colour is the result of simultaneously charging a slow bank power supply along with a fast bank.
When the modified LVF perturbation is applied at ∼20 ms into the discharge, it showed an out-board shift in radial plasma position as well as suppression of H α and O-I, C-III impurity line radiations. To confirm this observation, the LVF perturbation initiation time has been delayed by 10 ms in the next discharge, and the LVF pulse has been initiated at ∼30 ms into the discharge. The plasma parameters are compared for both discharges. Note that all other operational parameters are kept similar in both discharges. The comparison of time evolutions of plasma parameters in both discharges with the LVF perturbation applied at ∼20 ms (#25103) and ∼30 ms (#25102) is shown in figure 8.
In figure 8, the time traces for shot #25103 (LVF field applied at ∼20 ms) are shown with red color traces whereas the time traces for shot #25102 (LVF field applied at ∼30 ms) are shown in black color. The operating loop voltage (figure 8(a)) and pre-filled gas pressure are kept similar for both discharges. Figure 8(i) shows the applied LVF current pulses of similar magnitude (I max LVF ≈ 1.7 kA) in both discharges, corresponding to ∼100 G of the perturbation field. When the LVF perturbation field is delayed by 10 ms (applied at ∼30 ms instead of ∼20 ms into the discharge), the outward shift of the plasma position also got delayed by 10 ms (figure 8(f )). In this case, the H α (figures 8(c) and C-III (8(d)), O-I (8(e)) impurity line radiations, and REs dominated plasma current ( figure 8(b)) keep on rising during 20 ms to 30 ms time. This also reflects in the HXRs comparison in figure 8(h). The expanded plots of HXRs intensity and LVF current pulse during the period when the LVF current pulse is applied, are shown in figures 8(j) and (k) for LVF initiation time of ∼20 ms and ∼30 ms respectively. Note here that there is a slight inward convection in shot #25102 (black colored time traces) which is due to a slightly over-valued pre-fixed vertical magnetic field at that time instant. Shot #25103 would have had the  same inward advection at that time instant if the LVF pulse application would not have initiated at ∼20 ms. The time evolution of the chord average electron density is presented in figure 8(g), which clearly shows that the chord average electron density is identical for both discharges. This comparison indicates the role of LVF perturbation on plasma parameters improvement.

Summary
A very simple technique of in-house developed LVF perturbation is described in this paper to extract the REs from ADITYA tokamak plasma. A major design requirement for fabricating the new LVF coils is that they are capable of producing the perturbation field of the order of 150 G-300 G (at the plasma center, R =75 cm, vessel mid-plane (Z = 0)) calculated based on ADITYA plasma parameters. The LVF coils are designed, developed, commissioned, and tested successfully on ADITYA without involving any new or complicated technology. The experimental measurements of the localized vertical magnetic field outside the vessel showed a reasonable agreement with the simulation performed using COMSOL Multiphysics, commercially available software. The preliminary results of the start-up phase RE control reveal a significant reduction, almost about 80% reduction in the startup phase HXRs are achieved with the perturbation field of the order of ≈260 G-280 G when applied at ≈1 ms at the time of plasma initiation. It does not disrupt the plasma and extracts the REs without affecting the thermal component of the plasma. We observe a critical threshold of BLVF BT ⩾ 10 −2 is required to successfully mitigate the REs. This experimentally observed critical threshold is in good agreement with the modeling RE loss effectively using the LVF perturbation. Thus, the LVF perturbation technique is observed to be more reliable for REs extraction even when the gas-puff technique will not be helpful to control the runaways due to the higher impurities and lower temperatures of the plasma. Later, the LVF current pulsed has been modified by adding another capacitor bank to stretch the width of the current pulse, when applied at 20 ms of discharge length showed (i) radial movement of the plasma position at the low-field side and holds its study for the duration of ≈50 ms time from where the LVF pulse is applied (ii) Observed suppression in H α and O-I, C-III, visible continuum radiation (iii) observed reduction in REs-dominated plasma current as well as in HXRs (iv) observed plasma parameters improvement with better discharges quality as compared to discharges without LVF pulse.
REs mitigation using LVF perturbations may be envisaged as an additional tool for REs mitigation in bigger tokamaks including the ITER. Although the extrapolation of this technique for large-size machines including ITER needs more experimentation, this technique can be used in bigger tokamaks. Taking the design parameters of ITER (Toroidal field B T = 5.3 T, I p = 15 MA), based on the experimental results presented in this paper, vertical perturbation field (LVF) of ∼3%-4% of toroidal magnetic field i.e. ∼2-2.5 kG would be required at the plasma center in ITER. For the mitigation of disruption-generated REs, the degradation of the main plasma due to the perturbation field is not of concern. Hence, longer pulse durations of the LVF perturbations may also be used. Furthermore, after the disruption, the plasma becomes more resistive and will facilitate LVF penetration into the plasma. Therefore, the reported technique of REs mitigation through the application of a local vertical field is quite feasible in bigger tokamaks including ITER. This technique can very well be used in the real-time feedback mode as the LVF coil currents may be initiated at any desired time instant depending upon the generation time of the disruption-generated REs. The REs mitigation during the disruption phase with the application of LVF perturbation in real-time feedback mode has been attempted in ADITYA [34] with encouraging results. Note here that the RE suppression experiments with LVF in ADITYA tokamak are performed in a limiter configuration with a single poloidal ring limiter at one toroidal location. These experiments are successfully repeated in ADITYA Upgrade (ADITYA-U) tokamak, which has a toroidal belt limiter [47]. This shows that this technique remains independent of the plasma boundary and can be used in divertor configurations as well. In future work, we would like to carry out the experiments in ADITYA-U to mitigate REs by the application of the LVF perturbation during the start-up phase, flat-top, and disruption phases in a shaped-plasmas configuration using open divertor.