RFX-mod wall conditioning by lithium pellet injection

Munaretto, S.; Bello, S. Dal; Innocente, P.; Agostini, M.; Auriemma, F.; Barison, S.; Canton, A.; Carraro, L.; Fiameni, S.; Scarin, P.; Terranova, D.

doi:10.1088/0029-5515/52/2/023012

Cited by 16 publications

(13 citation statements)

References 39 publications

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“…Lithium-based surfaces are now routinely used in numerous magnetically confined fusion devices around the world. 1- 6 The study of lithium as a plasma-facing surface (PFS) had its beginnings with the work of Erents and McCracken in the early 1970s. 7 However, the first result demonstrating direct correlation between lithium-based surfaces and the behavior of the core fusion plasma was Mansfield's application of lithium coatings on the hot graphite limiter in the Tokamak Fusion Test Reactor (TFTR).…”

Section: Introductionmentioning

confidence: 99%

Lithium-based surfaces controlling fusion plasma behavior at the plasma-material interface

Allain¹,

Taylor²

2012

Physics of Plasmas

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The plasma-material interface and its impact on the performance of magnetically confined thermonuclear fusion plasmas are considered to be one of the key scientific gaps in the realization of nuclear fusion power. At this interface, high particle and heat flux from the fusion plasma can limit the material's lifetime and reliability and therefore hinder operation of the fusion device. Lithium-based surfaces are now being used in major magnetic confinement fusion devices and have observed profound effects on plasma performance including enhanced confinement, suppression and control of edge localized modes (ELM), lower hydrogen recycling and impurity suppression. The critical spatial scale length of deuterium and helium particle interactions in lithium ranges between 5-100 nm depending on the incident particle energies at the edge and magnetic configuration. Lithium-based surfaces also range from liquid state to solid lithium coatings on a variety of substrates (e.g., graphite, stainless steel, refractory metal W/Mo/etc., or porous metal structures). Temperature-dependent effects from lithium-based surfaces as plasma facing components (PFC) include magnetohydrodynamic (MHD) instability issues related to liquid lithium, surface impurity, and deuterium retention issues, and anomalous physical sputtering increase at temperatures above lithium's melting point. The paper discusses the viability of lithiumbased surfaces in future burning-plasma environments such as those found in ITER and DEMO-like fusion reactor devices. V C 2012 American Institute of Physics. [http://dx.

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Section: Introductionmentioning

confidence: 99%

Lithium-based surfaces controlling fusion plasma behavior at the plasma-material interface

Allain¹,

Taylor²

2012

Physics of Plasmas

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“…The EAST tokamak also recently obtained its first H-mode plasma after wall conditioning by lithium (Li) evaporation before plasma breakdown and the real-time injection of fine Li powder into the plasma edge [14]. Similar beneficial effects from lithium have been observed in other tokamaks including CPD [20], HT-7 [13], T-10 and T-11M [17], and the reversed field pinch device RFX [19]. The benefit of lithium applications on the plasma confinement and performance has been therefore demonstrated on a number of plasma devices and with various plasma configurations.…”

Section: Improved Plasma Confinement and Performance With Lithiummentioning

confidence: 76%

“…The NSTX lithium dropper has also been used successfully on the long-pulse superconducting EAST tokamak [14]. The application of lithium to PFC surfaces by the dropper technique is similar in principle to the lithium pellet injection technique used in the earlier TFTR and NSTX experiments, and more recently, on FRX [19] and the DOLLOP system in TFTR as described in Sec. 2.1.…”

Section: Lithium Delivery Systemsmentioning

confidence: 99%

“…The stellarator has the advantage of disruption-free operation. RFX [19] is a reversed field pinch (RFP) configuration which operates at a much higher plasma current for a given toroidal magnetic field value than tokamaks. Of the tokamak devices, CDX-U/LTX, CPD, and NSTX are of the low-aspect-ratio (R 0 /a ≤ 2) tokamak configuration, which is also called a "spherical tokamak" (ST) because of its spherical plasma shape.…”

Section: Lithium Experiments In Present Day Fusion Devices (2000mentioning

confidence: 99%

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Lithium As Plasma Facing Component for Magnetic Fusion Research

Ono¹

2012

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The use of lithium in magnetic fusion confinement experiments started in the 1990's in order to improve tokamak plasma performance as a low-recycling plasma-facing component (PFC). Lithium is the lightest alkali metal and it is highly chemically reactive with relevant ion species in fusion plasmas including hydrogen, deuterium, tritium, carbon, and oxygen. Because of the reactive properties, lithium can provide strong pumping for those ions. It was indeed a spectacular success in TFTR where a very small amount (~ 0.02 gram) of lithium coating of the PFCs resulted in the fusion power output to improve by nearly a factor of two. The plasma confinement also improved by a factor of two. This success was attributed to the reduced recycling of cold gas surrounding the fusion plasma due to highly reactive lithium on the wall. The plasma confinement and performance improvements have since been confirmed in a large number of fusion devices with various magnetic configurations including CDX-U/LTX (US), CPD (Japan), HT-7 (China), EAST (China), FTU (Italy), NSTX (US), T-10, T-11M (Russia), TJ-II (Spain), and RFX (Italy). Additionally, lithium was shown to broaden the plasma pressure profile in NSTX, which is advantageous in achieving high performance H-mode operation for tokamak reactors. It is also noted that even with significant applications (up to 1,000 grams in NSTX) of lithium on PFCs, very little contamination (< 0.1%) of lithium fraction in main fusion plasma core was observed even during high confinement modes. The lithium therefore appears to be a highly desirable material to be used as a plasma PFC material from the magnetic fusion plasma performance and operational point of view. An exciting development in recent years is the growing realization of lithium as a potential solution to solve the exceptionally challenging need to handle the fusion reactor divertor heat flux, which could reach 60 MW/m 2. By placing the liquid lithium (LL) surface in the path of the main divertor heat flux (divertor strike point), the lithium is evaporated from the surface. The evaporated lithium is quickly ionized by the plasma and the ionized lithium ions can provide a strongly radiative layer of plasma ("radiative mantle"), thus could significantly reduce the heat flux to the divertor strike point surfaces, thus protecting the divertor surface. The protective effects of LL have been observed in many experiments and test stands. As a possible reactor divertor candidate, a closed LL divertor system is described. Finally, it is noted that the lithium applications as a PFC can be quite flexible and broad. The lithium application should be quite compatible with various divertor configurations, and it can be also applied to protecting the presently envisioned tungsten based solid PFC surfaces such as the ones for ITER. Lithium based PFCs therefore have the exciting prospect of providing a cost effective

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“…To improve density control and to contribute to the worldwide research on plasma facing components, a lithization program has been recently started, in addition to more traditional wall-conditioning techniques such as Glow Discharge Cleaning and Boronization [4]. Lithium wall conditioning has been applied by evaporation with a Liquid Lithium Limiter [6] and with single pellet injection [7]. Pellets injected about an overall amount of 1 g of Li, while the total amount of evaporated Li in the first campaign is about 14 g. Significant improvement of the density control is observed.…”

Section: Plasma-wall Interaction and Density Controlmentioning

confidence: 99%

Overview of the RFX-mod fusion science programme

et al. 2013

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This paper reports the highlights of the RFX-mod fusion science programme since the last 2010 IAEA Fusion Energy Conference. The RFX-mod fusion science programme focused on two main goals: exploring the fusion potential of the reversed field pinch (RFP) magnetic configuration and contributing to the solution of key science and technology problems in the roadmap to ITER. Active control of several plasma parameters has been a key tool in this endeavour. New upgrades on the system for active control of magnetohydrodynamic (MHD) stability are underway and will be presented in this paper. Unique among the existing fusion devices, RFX-mod has been operated both as an RFP and as a tokamak. The latter operation has allowed the exploration of edge safety factor qedge < 2 with active control of MHD stability and studies concerning basic energy and flow transport mechanisms. Strong interaction has continued with the stellarator community in particular on the physics of helical states and on three-dimensional codes.

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RFX-mod wall conditioning by lithium pellet injection

Cited by 16 publications

References 39 publications

Lithium-based surfaces controlling fusion plasma behavior at the plasma-material interface

Lithium-based surfaces controlling fusion plasma behavior at the plasma-material interface

Lithium As Plasma Facing Component for Magnetic Fusion Research

Overview of the RFX-mod fusion science programme

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