## 1 Introduction  
Maintaining an ultra-high vacuum environment in a magnetic confinement fusion device is a fundamental condition for fusion plasma start-up and maintenance. Vacuum leaks may result from design defects or mechanical breakdowns in the operation of vacuum bellows [1] and other components. Once a vacuum failure occurs, it will be necessary to vent the vacuum vessel for the repair or replacement of broken components. Additionally, after the re-vacuuming, wall conditioning is performed to remove residue gas on the wall, reduce edge recycling, and recover well-confined plasma. For the plasma recovery of HT-7 (Hefei Tokamak-7), boronization is conducted to suppress impurities, and the ion cyclotron range of frequency (ICRF)-cleaning discharges are operated for hydrogen removal and wall condition improvement [2]. The vessel conditioning methods of the Joint European Torus (JET) include baking and glow discharge cleaning (GDC), which is used to remove water and gas impurities, and beryllium evaporation is carried out to reduce oxygen and carbon concentrations [3,4]. After the vacuum leak of Tore Supra, baking, GDC, and ICRF discharges were conducted to remove hydrogen and other absorbed impurities [5]. In EAST, after repairing the vacuum failure, normally, several wall conditioning techniques would also be conducted successively, i.e., baking, \(\mathrm{D_{2}}/\mathrm{He}\)-GDC, and ICRF-DC [6]. When the vacuum chamber of the device is baked and cleaned for enough time, a process of plasma recovery is started. The main purpose in this plasma recovery process is to achieve well-confined plasma maintenance throughout the whole discharge. During this process, the establishment and maintenance of steady-state, well-confined plasma would be rather difficult without the application of additional wall conditioning techniques (e.g., GDC or lithium evaporation). Normally, there are 1-2 weeks of runaway discharges in which the plasma filled with high-energy runaway electrons--so called \"non-collisional plasma\", with a very low electron temperature--and the plasma current might be carried by runaway electrons [7]. It was found that the impurity behavior plays an important role in plasma recovery. High levels of impurities such as oxygen, nitrogen, carbon dioxide, and water are present in the vacuum chamber, making it difficult to maintain well-confined plasma and increase electron temperature [8]. Usually, low-Z material coating (evaporation), such as lithium [9,10] or boron, is routinely utilized to enhance the wall condition and impurity suppression, and finally, accomplish the plasma recovery. Once the confined plasma can be stably and repetitively sustained, the experimental campaign moves on to the next phase, which is normal plasma operation or the physical experimental phase. The results from several fusion devices have proved a positive effect of lithium coating and lithium granule injection for impurity control and plasma performance in this phase, e.g.,  
1. a lower tungsten density in the core plasma, lower tungsten influx, and the achievement of a higher energy confinement time and density limit in T-10 tokamak [11];
2. a reduced peak heat flux on the divertor from 5 to 2 MW m\({}^{-2}\), with 150 and 300 mg of pre-discharge lithium evaporation in NSTX [12];
3. an increased duration of enhanced H-mode pedestal phases of up to 350 ms with the injection of a powder of lithium particles 45 um in diameter in DIII-D [13];
4. long-pulse, high-confinement H-mode discharge with a duration of over 30 s through advanced lithium wall conditioning in EAST [14]. It is important to shorten the plasma recovery time for the existing devices after vacuum failure occurs, and this is also essential to allow future devices such as ITER and CFETR to quickly obtain the first plasma. Unfortunately, many diagnostics are not available in plasma recovery due to the unsuitable operation conditions involved and the short lifetime of confined plasma in the discharges.