Application of lithium in systems of fusion reactors. 2. The issues of practical use of lithium in experimental facilities and fusion devices

Apicella M.L., Lazarev V., Lyublinski I., Mazzitelli G., Mirnov S., Vertkov A.

Liquid lithium use on the base of capillary porous systems (CPS) application as plasma facing material (PFM) of tokamaks is advanced way to solve the problems of plasma contamination with high Z impurity, PFM degradation and tritium retention. In frame of joint program between ENEA (Italy) and FSUE ‘Red Star’ & TRINITI (RF) started at the end of 2005 the test of passive cooled liquid lithium limiter (LLL) with CPS in a high field, medium size, carbon free tokamak FTU have been performed successfully. The LLL has been inserted in ohmic plasma discharges and at additional heating with LH and ECR at power levels in the MW range without any particular problem ( B T  = 6 T, I p  = 0.5–0.9 MA, n e  = 0.2–2.6 × 10 20  m −3 , τ  = 1.5 s, P ∼2–5 MW/m 2 at a normal discharge). The behavior of lithium CPS based on stainless steel wire mesh and its surface modification in normal discharges and at disruptions has been studied. Results of microscopic analyses of CPS structure after experimental campaigns are presented. The possibility to withstand heat load exceeding 5 MW/m 2 without structural damage, lithium surface renewal, mechanical stabilization of liquid lithium against MHD forces have been confirmed. Application of W, Mo as the base material and possible structure types of CPS have been considered for operating parameters improvement of long-living plasma facing components.

Lyublinski I.E., Vertkov A.V., Evtikhin V.A.

A short analysis of lithium properties presented in this article was carried out in terms of lithium application in various systems of a fusion reactor. The article is intended for use by the engineers and researchers involved in activities on thermonuclear fusion.

Vertkov A., Luyblinski I., Evtikhin V., Mazzitelli G., Apicella M.L., Lazarev V., Alekseyev A., Khomyakov S.

A liquid lithium limiter (LLL) with a total area of 170 cm 2 and lithium amount of ≅80 g based on lithium filled capillary-pore system (CPS) as an advanced plasma facing material has been tested for the first time on the medium size high magnetic field tokamak FTU. The main technological aims of the experiment were to test: the compatibility of lithium CPS in real plasma tokamak conditions; the capacity of CPS to self-regenerate an exposed surface and to confine lithium during normal plasma operations and disruptions; the capability to withstand high heat loads without damage of limiter surface. Lithium CPS advantages, experimental base for the tokamak application, general description of LLL design and preparation method were described. No anomalous phenomenon like “lithium bloom” has occurred in the plasma and no damage of CPS surface and LLL structure has been observed after ∼60 discharges with power flux up to 5 MW/m 2 , as well as on plasma disruptions. Self-restoring of lithium surface has been confirmed. LLL application as a conditioning element with depositing a lithium film on the walls (litization) has resulted in radiation losses, plasma contamination and D 2 recycling reduction. Perspectives of lithium CPS application for further fusion reactors have been considered.

Mirnov S.V., Azizov E.A., Evtikhin V.A., Lazarev V.B., Lyublinski I.E., Vertkov A.V., Prokhorov D.Y.

The paper is an overview of recent results of Li limiter testing in T-11M tokamak. The lithium limiter is based on the capillary-pore system (CPS) concept. The Li erosion process and deuterium (D2) and helium (He) sorption by Li first wall were investigated. The ability of capillary forces to confine the liquid Li in the CPS limiter during disruption was demonstrated. The idea of combined lithium limiter with thin (0.6 mm) CPS coating as a solution of the heat removal problem was realized. As a result the quasi steady-state tokamak regime with duration up to 0.3 s and clean (Z eff = 1) deuterium plasma has been achieved. The temporal evolution of the lithium surface temperature during discharge was measured by a IR radiometer and then was recalculated to the surface power load. For the estimation of the Li limiter erosion the Li neutral and ions spectral line emission were observed. The increase in lithium erosion as a result of limiter heating was discovered. The radial distribution of plasma column radiation measurements showed up to 90% of the total radiation losses in a relatively thin (5 cm) boundary layer and only 10% in a plasma centre during discharges with high Li influx. Oscillations of Li emission and saw-tooth-like oscillations of the limiter surface temperature have been detected in discharge regimes with highest Li limiter temperature (>600 °C). A version of Li CPS first wall of DEMO reactor and Li CPS limiter experiment in the International Thermonuclear Energy Reactor are suggested.

Lyublinski I.E., Evtikhin V.A., Vertkov A.V., Ezhov N.I., Shcherbakov V.M.

MHD pressure drop in flowing liquid metal for a tokamak with high magnetic fields is a key concern regarding the development of lithium self-cooled test module for ITER and lithium breeding blanket for DEMO-type projects. MHD losses on liquid metal pumping can be most efficiently reduced by applying of an electrically insulating coating to the inner surface of the channels. The developed method to evaluate the electrically insulating properties of a coating placed on the internal surface of a vanadium test channel is based on the pressure drop measurement in a liquid lithium forced circulation system with the test section in magnetic field. The tests were conducted on channels of V–4Ti–4Cr alloy with and without on insulating coating based on the aluminum nitride technique (AlN) [1] . The dimensions of the test section were 6 mm × 20 mm × 320 mm. The tests involved lithium flow velocities up to 3 m/s, temperatures up to 400 °C and a uniform transverse magnetic field up to 1.6 T. The measured hydraulic resistances have shown five times reduction for the coated wall channel in comparison with the conducting wall channel.

Apicella M.L., Mazzitelli G., Lazarev V.B., Azizov E.A., Mirnov S.V., Petrov V.G., Evtikhin V.A., Lyublinski I.E., Vertkov A.V., Lucca F., Ferdinando E.D., Mazzone G., Ramogida G., Roccella M.

The possibility of using liquid lithium as plasma facing component for the divertor target plates will be investigated on FTU by employing an innovative solution that exploits the surface tension forces in capillary channels to compensate j  ×  B forces induced in lithium. For this structure, a high stability and resistance is foreseen as well as an intrinsic capability to self-regenerate the plasma facing surface. FTU represents the first very good opportunity to test this configuration, called capillary porous system (CPS), in an ITER relevant experiment at high plasma density (up to 3.2 × 10 20  m −3 ), high current (up to 1.6 MA) and high magnetic field (up to 8 T). As first step of its application on FTU, foreseen for the second half of 2005, CPS system will be employed for a “litization” experiment aimed to reduce plasma contamination and recycling.

Federici G., Zhitlukhin A., Arkhipov N., Giniyatulin R., Klimov N., Landman I., Podkovyrov V., Safronov V., Loarte A., Merola M.

This paper describes the response of plasma facing components manufactured with tungsten (macro-brush) and CFC to energy loads characteristic of Type I ELMs and disruptions in ITER, in experiments conducted (under an EU/RF collaboration) in two plasma guns (QSPA and MK-200UG) at the TRINITI institute. Targets were exposed to a series of repetitive pulses in QSPA with heat loads in a range of 1-2 MJ/m 2 lasting 0.5 ms. Moderate tungsten erosion, of less than 0.2 μm per pulse, was found for loads of ∼1.5 MJ/m 2 , consistent with ELM erosion being determined by tungsten evaporation and not by melt layer displacement. At energy densities of ∼1.8 MJ/m 2 a sharp growth of tungsten erosion was measured together with intense droplet ejection. MK-200UG experiments were focused on studying mainly vapor plasma production and impurity transport during ELMs. The conditions for removal of thin metal deposits from a carbon substrate were characterized.

Evtikhin V.A., Lyublinski I.E., Vertkov A.V., Azizov E.A., Mirnov S.V., Lazarev V.B., Sotnikov S.M., Safronov V.M., Prokhorov A.S., Korzhavin V.M.

At present the most promising principal solution of the divertor problem appears to be the use of liquid metals and primarily of lithium Capillary-Pore Systems (CPS) as of plasma facing materials. A solid CPS filled with liquid lithium will have a high resistance to surface and volume damage because of neutron radiation effects, melting, splashing and thermal stress-induced cracking in steady state and during plasma transitions to provide the normal operation of divertor target plates and first-wall protecting elements. These materials will not be the sources of impurities inducing an increase of Zeff and they will not be collected as dust in the divertor area and in ducts. Experiments with lithium CPS under simulating conditions of plasma disruption on a hydrogen plasma accelerator MK-200 [~(10 - 15) MJ/m2, ~50 μs] have been performed. The formation of a shielding layer of lithium plasma and the high stability of these systems have been shown. The new lithium limiter tests on an up-graded T-11M tokamak (plasma current up to 100 kA, pulse length ~0.3 s) have been performed. Sorption and desorption of plasma-forming gas, lithium emission into discharge, lithium erosion, deposited power of the limiter are investigated in these experiments. The first results of experiments are presented.

Federici G., Andrew P., Barabaschi P., Brooks J., Doerner R., Geier A., Herrmann A., Janeschitz G., Krieger K., Kukushkin A., Loarte A., Neu R., Saibene G., Shimada M., Strohmayer G., et. al.

Some of the remaining crucial plasma edge physics and plasma–material interaction issues of the ITER tokamak are discussed in this paper, using either modelling or projections of experimental results from existing tokamak operation or relevant laboratory simulations. The paper covers the following subject areas at issue in the design of the ITER device: (1) plasma thermal loads during Type I ELMs and disruptions, ensuing erosion effects and prospects for mitigating measures, (2) control of co-deposited tritium inventory when carbon is used even on small areas in the divertor near the strike points, (3) efficiency of edge and core fuelling for expected pedestal densities in ITER, and (4) erosion and impurity transport with a full tungsten divertor. Directions and priorities of future research are proposed to narrow remaining uncertainties in the above areas.

Smith D.L., Park J.-., Natesan K.

A key issue for the self-cooled lithium blanket concept with a vanadium alloy structure is the development of an electrically insulating coating on the coolant channel walls to mitigate the magneto-hydrodynamic pressure drop in a high magnetic field. A systematic investigation of the thermodynamics and kinetics of oxygen and calcium interactions in the vanadium alloy/lithium system is being conducted to define the system parameters required for in situ formation of a CaO coating on vanadium alloys. This paper presents results of theory and modeling as well as experimental results on the formation of CaO coatings on vanadium alloys after exposure at temperatures of 600–700 °C to lithium with a small fraction of Ca added. Coatings of 10–30 μm with high electrical resistivity (>108 Ω  cm) have been formed on V-alloys.

Smith D.L., Park J.-., Lyublinski I., Evtikhin V., Perujo A., Glassbrenner H., Terai T., Zinkle S.

Development of effective and reliable coatings is a key to the viability of most, if not all, fusion blanket systems. The specific purpose and requirements of the coatings vary widely, depending on the blanket concept. The efforts on coating development to date have focused primarily on electrically insulating coatings for the self-cooled lithium concepts with a vanadium alloy structure, and the tritium barrier coatings for the water-cooled, Pb–Li (WCLL) breeder concepts with ferritic steel structures. Although other coating materials are under consideration, most of the effort on the electrically insulating coatings has focused on CaO and AlN coatings on V–4Cr–4Ti alloy structure. Most of the effort on the tritium barrier coating development is focused on Al2O3 coatings formed on aluminized ferritic steels. This paper presents an overview of the status of coating development for the various fusion concepts with emphasis on the materials interaction and chemistry control issues associated with the formation, stability and performance of the coatings.

Evtikhin V.A., Lyublinski I.E., Vertkov A.V., Mirnov S.V., Lazarev V.B., Petrova N.P., Sotnikov S.M., Chernobai A.P., Khripunov B.I., Petrov V.B., Prokhorov D.Y., Korzhavin V.M.

The ITER project development has shown that considerable difficulties are encountered when already known engineering solutions and materials are used for divertor and divertor plates for tokamaks of such a scale. We offer to use a Li capillary-pore system (CPS) as a plasma facing material for tokamak divertor. Evaporated Li serves as a gas target and redistributes thermal load. The heat flux from the plasma is transported to the first wall by Li radiation in the plasma periphery. This allows the divertor plate to reduce the heat flux. A solid CPS filled with liquid lithium has a high resistance to surface damage in the stationary mode and during plasma transitions (disruptions, ELMs, VDEs, runaways) to assure normal operation of the divertor target plates. These materials are not the sources of impurities giving rise to Zeff and they will not be collected as dust in the divertor area and in ducts. Experiments with lithium CPS in a steady-state mode (up to 25 MW m-2) and in plasma disruption simulation conditions (~5 MJ m-2, ~0.5 ms) have been performed. High stability of these systems have been shown. Li limiter tests on T-11M tokamak have revealed the lithium CPS compatibility with the edge plasma for energy loads of up to 10 MW m-2. In a stable discharge mode at lithium limiter temperature of 20-600°C, no Li abnormal erosion and injection to plasma have been detected. A high sorption of D+ and H+ ions on the vessel walls was the main substantial result of the replacement of a graphite limiter by lithium one. He and D sorption was terminated by wall heating up to 50-100°C and above 350°C, respectively. T-11 tests on helium discharge allowed to reduce limiter heat load by a factor of two due to lithium radiation. All the experimental results have shown considerable progress in the development of lithium divertor.

Vertkov A.V., Evtikhin V.A., Lyublinski I.E.

The existing technological approaches for the formation of nitride- and oxide-based self-healing electrical insulating coatings for vanadium alloy–lithium systems are considered. The results of the property study of coatings applied from liquid lithium containing Al, N, Si, B additions on various modes are considered. The formation conditions of AlN-based coatings with scale specific electrical resistivity (∼50 Ω m) on the V–4Ti–4Cr vanadium alloy are determined. The results of formation and stability research of coatings on the V–4Ti–4Cr vanadium alloy in convectional and forced circulating lithium with Al and N additions in the homogeneous and heterogeneous lithium systems are discussed.

Evtikhin V.A., Lyublinski I.E., Vertkov A.V., Yezhov N.I., Khripunov B.I., Sotnikov S.M., Mirnov S.V., Petrov V.B.

Experimental results of complex studies of lithium capillary-pore systems (CPS) for application as a plasma facing structure in divertor and on the first wall of a fusion reactor are reported. The ability of CPS to accept and to remove high heat fluxes (up to 30 MW m −2 ) in steady-state conditions (tens of minutes) has been evaluated on target plate imitator mock-ups supplied with cooling and lithium feed systems under electron beam power load in a linear plasma facility. Experimental study of lithium flow up to 2.5 m s −1 in CPS made of material with final conductivity for various mesh sizes and of the effect of cross magnetic field up to 1.6 T on its parameters has been made. The results of successful experiments on the T-11M tokamak helium and hydrogen plasma interaction with a CPS-based lithium limiter and lithium puff influence on the plasma performances are presented and analysed.

Evtikhin V.A., Lyublinski I.E., Vertkov A.V., Belan V.G., Konkashbaev I.K., Nikandrov L.B.

Estimations and experimental studies on the possibility of capillary-pore systems with lithium as the first wall and divertor target plate plasma-facing material have been conducted in support of the lithium liquid metal fusion reactor concept. The possibility of the lithium-filled capillary-pore systems to withstand a high energy plasma flux has been demonstrated.

Parsons M.S., Porcelli M., Emdee E.D., Goldston R.J.

Abstract The lithium vapor divertor concept is being developed as a method to achieve detached divertor conditions in a tokamak while minimizing impurity radiation losses from the core plasma. SOLPS-ITER modeling has previously been used to identify some of the geometric constraints and required lithium evaporation rate of a lithium vapor divertor in a medium-sized tokamak during steady-state operation. Here an updated conceptual design based on these operating requirements is introduced and the thermal response of the system is modeled during cyclical operation, consistent with operation in a short-pulse tokamak. Controllability of the temperature of the lithium capillary porous system (CPS) is achieved by adopting a design where there is no line-of-sight for radiation from the plasma to reach the heated CPS surface. Operational strategies to minimize the amount of lithium evaporated between plasma discharges while achieving steady evaporation rates during plasma discharges are discussed and modeled here. The optimal feedforward control strategy demonstrated in this work is to ramp up the temperature of the evaporator as quickly as possible immediately before a plasma discharge and then reduce the heating to match the desired steady-state net evaporation rate just before the plasma discharge begins, allowing the thermal inertia of the system to stabilize the evaporation rate during the first second of the plasma discharge.

Zhang D.H., Meng X.C., Zuo G.Z., Li X., Zhang J., Zhang X.X., Yan Q.Z., Cao B., Hu J.S.

López Pérez C., Marchhart T., Marin A., Nieto Perez M., Allain J.P.

Liquid lithium wettability has been investigated on novel zirconium-alloyed porous tungsten fabricated by the place-holder spark plasma sintering technique. This paper investigates the wetting properties of liquid lithium on the substrates by in vacuo contact angle measurements of lithium droplets at Penn State’s IGNIS-2 facility. X-ray Photoelectron Spectroscopy analysis was performed on the samples to investigate the role of the surface chemistry in the contact angle behavior of lithium on the substrates. It was seen that hydrated and hydroxylated states of the surface play a key role in affecting the lithium affinity of the surface.

D'Ovidio G., Martín-Fuertes F., Marugán J.C., Bermejo S., Nitti F.S.

Several experimental facilities, mainly focused on nuclear fusion applications, propose lithium as main coolant and/or tritium breeder material, limiter or divertor for its advantageous properties. Lithium fire hazard represents a critical risk for the production of an important population of reactive, corrosive and toxic aerosols, and for the potential mobilization, transport and release of radioactive species, initially retained in the molten metal, as in experimental facilities like the IFMIF-DONES accelerator neutron source. Consequently, a specific methodology should be developed for managing possible lithium fire scenarios that could occur during the lifetime of these unique facilities. Applying the Defense-in-Depth principle to minimize the fire risk in the particular case of the IFMIF-DONES plant, this work describes a set of passive and active measures for lithium fire prevention, detection and mitigation in compliance with main international and national standards on fire protection. According to the present safety analyses, active measures, including the use of fixed extinguishing systems employing chemical agents, do not seem to be entirely reliable in case of large lithium fires, for which more preferable passive fire protection measures are also being considered to be implemented in the final design of IFMIF-DONES, such as room inertization, catch pans and drain systems, constituting together a main line of defense to safely manage potential lithium fire scenarios.

Mavrin A.A., Pshenov A.A.

An 0D model is proposed that makes it possible to estimate the limiting stationary heat loads to the targets covered with liquid lithium (LL) layer, taking into account the effects of vapor shielding by sputtered and evaporated LL and hydrogen recycling. Several models of cooled target substrates are considered in which the LL layer facing the plasma is placed. For the considered substrate models, a parametric analysis of the tolerable stationary heat loads to the target on the substrate thickness, the effective cooling energy per particle of sputtered lithium, and the lithium prompt redeposition factor was carried out. It is shown that, at a small substrate thickness, the choice of the substrate model has a significant impact on the tolerable heat loads. It is also shown that even at unrealistically large values of the effective cooling energy, the dissipation of lithium remains modest. This means that in regimes with a high power coming from the core plasma to the edge, the injection of an additional radiator is required. Finally, it is shown that one of the most effective ways to increase the tolerable stationary heat loads would be to reduce the thickness of the target substrate.

Liu J., Jing W., Guo H., Gao Y., Wang S., Chen B., Chen J., Wang H., Wei J., Ye Z., Gou F.

In this paper, an embedded multichannel capillary porous system (EM-CPS) was designed and fabricated with 304 stainless steel using the laser ablation method. The EM-CPS revealed its excellent ability to wick liquid lithium to its surface effectively. The interaction between Li-prefilled EM-CPS and plasma was studied, and the results showed that the surface temperature decreased by ~140 °C compared with the results of the experiment of EM-CPS without lithium filling. Additionally, EM-CPS displayed a better heat transfer performance and stronger radiation loss of the vapor cloud than the traditional woven tungsten-based meshes. In addition, the drift of the lithium vapor cloud center was found during plasma irradiation and led to a decrease in the intensity of the Li 670.78 nm emission line detected by the spectrometer at the observation point. When the thermal load deposited on the sample surface is reinforced by increasing the magnetic field, the rise in surface temperature is restrained due to the enhanced heat dissipation capability of lithium. SEM images of irradiated samples showed that the 304 stainless steel-based EM-CPS has corrosion problems due to the interaction between liquid lithium and argon plasma, but it still showed good plasma-facing characteristics. These findings provide a reference for further studies of embedded multichannel CPSs with plasma-facing components (PFCs) in linear plasma devices and tokamaks in the future.

Vertkov A.V., Zharkov M.Y., Lyublinskii I.E., Safronov V.A.

When developing the stationary fusion reactor, an unresolved issue is the design of its intra-chamber plasma-facing elements. It has now become obvious that among the materials conventionally used for intra-chamber elements, there are no solid structural materials that would meet the requirements for the long-term operation under the effect of the flux of fusion neutrons (14 MeV) with a density of ~1014 cm–2 s–1 and the heat flux with a power density of 10–20 MW/m2. An alternative solution to this problem is the use of liquid metals as a plasma-facing materials, and, first of all, the use of lithium, which has a low atomic number (low charge number Z). Other easily-melting metals are also considered, which have higher Z number, but lower saturation vapor pressure than lithium. This will make it possible to create the long-lived, heavy-to-damage and self-renewing surface of the intra-chamber elements, which will not contaminate the plasma. The main ideas of the alternative concept of the intra-chamber elements can be formulated based on the comprehensive analysis of the problems and requirements arising during the development of intra-chamber elements of the stationary reactor, for example, the DEMO-type reactor. The article presents the analysis of the possible design of the lithium-coated intra-chamber elements and discusses the main ideas of the lithium first wall concept for the tokamak with reactor technologies.

Rindt P., van den Eijnden J.L., Morgan T.W., Lopes Cardozo N.J.

• A realistic liquid metal divertor concept is designed for the European DEMO reactor. • An armor of 3D-printed porous tungsten, filled with liquid tin, is used. • Heat loading capability is increased compared to baseline designs, in both steady state operation and during slow transients. • Resilience against disruptions is not outside the realm of possibility, but requires experimental testing. Liquid metal (LM) divertors are considered for the European DEMO reactor, because they may offer improved performance compared to the tungsten monoblock concept. The goal of this work is to provide a concept design, and explore the limitations of liquid metal divertors. To this end, a set of design requirements was formulated in close collaboration with the EUROfusion Power Plant Physics and Technology team (responsible for the design of the EU-DEMO). Tin was chosen as the preferred liquid metal, because unacceptable Tritium retention issues arise when lithium is used in DEMO. A concept design was then chosen that consists of water cooled pipes that are square on the outside and round on the inside, a corrosion barrier, and a 3D-printed porous tungsten armor layer filled with liquid tin. The porous armor layer acts as a Capillary Porous System (CPS). The design was analyzed using thermo-mechanical FEM simulations for various armor thicknesses and heat sink materials: Densimet, W/Cu composites, and CuCrZr. The highest heat loading capability achieved is 26.5 MW/m 2 in steady state (18.9 MW/m 2 when taking into account a safety margin of 1.4). This is achieved using a CuCrZr pipe, with a 1.9 mm thick armor. When increasing the armor layer to 3 mm thick, more than 80 MW/m 2 can be withstood during slow transients thanks to vapor shielding, but at the same time the steady-state capability is reduced to 18 MW/m 2 . Resilience against disruptions cannot yet be proven, but is deemed within the realm of possibility based on estimates regarding the behavior of vapor shielding. This should be further investigated. Overall, the concept is considered a significant improvement compared to the original specifications (which are also the specifications to the tungsten monoblocks: 10 MW/m 2 in steady state, and ∼ 20 MW/m 2 during slow transients). Moreover, the possibility of withstanding disruptions is regarded as a potentially major improvement.

Garkusha I.E., Makhlai V.A., Petrov Y.V., Herashchenko S.S., Ladygina M.S., Aksenov N.N., Byrka O.V., Chebotarev V.V., Kulik N.V., Staltsov V.V., Pestchanyi S.

This paper presents experimental studies of plasma-surface interactions during powerful plasma impacts of a Quasi-Stationary Plasma Accelerator (QSPA) on the Sn capillary porous systems (CPSs) in conditions, simulating disruption and Edge Localized Modes (ELM) like loads. Experiments were carried out using two QSPA devices. ELM-like plasma exposures were performed with QSPA-M test-bed facility. A large-scale QSPA Kh-50 device was used to simulate plasma disruptions and giant ELMs. Variation of the plasma stream energy density has been performed to study the onset of vapour shield. It is shown that during plasma exposures of a Sn-CPS target with the QSPA plasma load 0.5 MJ/m2, which corresponds to the strong vapour shielding of the exposed surface. A comparison between the obtained results on the vapour shielding of Sn CPS and available numerical simulation using the TOKES code has been performed.

Nygren R.E., Youchison D.L., Michael J.R., Puskar J.D., Lutz T.J.

• Unexpected rapid failure of a small 1018 mild steel vessel preheated to ~200 °C and under vacuum occurred when lithium at ~400 °C and ~1 atm. flowed into the vessel. • Fractography confirms failure enabled by liquid metal embrittlement. In preparation for testing a lithium-helium heat exchanger at Sandia, unexpected rapid failure of the mild steel lithium preheater due to liquid metal embrittlement occurred when lithium at ~400 °C flowed into the preheater then at ~200 °C. This happened before the helium system was pressurized or heating with electron beams began. The paper presents an analysis of the preheater plus a discussion of some implications for fusion.

Rindt P., Korving S.Q., Morgan T.W., Lopes Cardozo N.J.

Abstract A fusion reactor divertor must withstand heat flux densities <10 MW m−2. Additionally, it may have to withstand millisecond pulses on the order of 0.5 to 30 MJ m−2 due to (mitigated) edge-localized modes (ELM) occurring with 30 to 60 Hz. We investigate if these requirements can be met by capillary porous system (CPS) liquid lithium divertors (LLD). 3D-printed tungsten CPS targets were exposed in the linear plasma device Magnum-PSI, to deuterium plasma discharges lasting 15 s, generating 1.5 to 16 MW m−2, and T e ∼ 1.5 eV. Additionally, ELM-like pulses were superimposed on top of the steady state for 3 s with a frequency of 2 and 100 Hz, power flux densities of 0.3 to 1 GW m−2, and T e up to ∼14 eV. All Li targets survived without damage. The surface temperature (T s) was locked at ∼850 °C, which was attributed to power dissipation via vapor shielding. Meanwhile, unfilled reference targets melted during the severest pulsed loading. A blue grayish layer of presumably LiD formed when T s < 500 °C, but disappeared when the locking temperature was reached. This implies that LiD formation can be avoided, but that it may require a surface temperature at which Li evaporation excessively contaminates the core plasma in a tokamak. During pulsed loading the plasma facing surface remained wetted in all conditions. Bolometry indicated that, only during pulses, there was a large increase in radiative power dissipation compared to targets without Li. A high speed camera with a Li-I filter showed that strong Li evaporation continued up to 5 ms after a pulse. Overall, the liquid-lithium-filled 3D-printed tungsten targets were found to be highly robust, and 3D-printing can be considered as a promising manufacturing technique for LLDs. Further research is needed particularly on the formation of LiD and the associated tritium retention, as well as the impact of enhanced evaporation during and after ELMs on the core plasma.

Gribkov V.A., Demina E.V., Demin A.S., Maslyaev S.A., Pimenov V.N., Prusakova M.D., Sirotinkin V.P., Rogozhkin S.V., Lyamkin P.V., Padukh M.

The effect of high-power pulsed fluxes of deuterium ions and deuterium plasma generated in the Plasma Focus PF-1000U device on oxide dispersion strengthened ferritic steel KP4-ODS (Fe–15 Cr–4 Al–2 W–0.35 Y2O3) was experimentally studied. When the samples were irradiated with two pulses (N = 2), the plasma flux power density was qpl ≈ 108 W/cm2 and that of ion beam qi × 109 W/cm2. At N = 9, qpl ≈ 2 ×108 W/cm2 and qi ≈ 5 × 109 W/cm2. The pulse duration of the plasma beams was τpl ≈ 100 ns and that of the ion beams τi ≈ 50 ns. It was shown that irradiation of the material in the soft mode (N = 2) leads to surface erosion due to evaporation of the material and is accompanied by the surface polishing effect. In this case, there is no significant change in the initial structural phase state of steel; only a small change in the crystal lattice parameters of solid solutions based on iron and chromium is observed. In the hard irradiation mode (N = 9), owing to the high heating of the surface layer, in addition to erosion, the material melts. In the structure of the surface layer of the ODS steel, a chromium-based solid solution disappears and only an iron-based solid solution remains, while the number of second-phase nanoparticles increases. The presence of a liquid phase formed upon exposure to fluxes of deuterium ions and deuterium plasma stimulates the possibility of complete dissolution of small (less than ~20 nm) nanoparticles of Y2O3 oxide and partial dissolution of larger (tens of nanometers) nanoparticles. Enhanced in comparison with the solid phase, the diffusion redistribution of elements in the molten surface layer contributes to the formation of Y2O3 nanoparticles and oxides of other elements that make up the ODS steel (Al2O3, Y–Al–O) upon cooling of the melt.

Andruczyk D., Maingi R., Kessel C., Curreli D., Kolemen E., Canik J., Pint B., Youchison D., Smolentsev S.

While high-Z solid plasma-facing components (PFCs) are the leading candidates for reactors, it is unclear that they can survive the intense plasma material interaction (PMI). Liquid metals (LM) PFCs offer potential solutions since they are not susceptible to the same type of damage, and can be “self-healing”. Following the Fusion Energy System Study on Liquid Metal Plasma Facing Components study that recently was completed by Kessel et al. (Fusion Sci Technnol 75:886, 2019) a domestic LM PFC design program has been initiated to develop reactor-relevant LM PFC concepts. This program seeks to evaluate LM PFC concepts for a Fusion Nuclear Science Facility (FNSF) or a Compact Pilot Plant via engineering design calculations, modeling of PMI and PFC components and laboratory experiments. The latter involves experiments in dedicated test stands and confinement devices and seeks to identify and answer open questions in LM PFC design. The new national LM PFC program is first investigating lithium as the plasma facing material for a flowing divertor PFC concept. Several flow speeds will be evaluated, ranging from ~ cm/s to m/s. The surface temperature will initially be held below the strongly evaporative limit in the first design; higher temperatures with strong evaporation will be considered in future concepts. Other topics of interest include: understanding of the hydrogen and helium interaction with the liquid lithium; single effect experiments on wetting, compatibility and embrittlement; and prototypical experiments for control and characterization of flowing LM. A path to plasma and future tokamak exposure of these concepts will be developed.

Alegre D., Oyarzabal E., Tafalla D., Liniers M., Soleto A., Tabarés F.L.

In a future fusion reactor like DEMOnstration reactor (DEMO) one of the main concerns is the handling of the power exhaust from the plasma, especially at the divertor. The expected power loads cannot easily be handled by traditional armor solutions based on solid materials like tungsten, especially when the effect of intense neutron bombardment is also considered. Interest in armor concepts based on liquid metals has been subsequently on the rise, as they prove to be more resilient against high, fast power loads and neutron bombardment. However, engineering solutions for those concepts are very complex, and need to be tested. For this purpose, Optimization of Liquid Metal Advanced Targets project (OLMAT) has been envisaged. The project will use the Neutral Beam Injection of the TJ-II stellarator to irradiate liquid metal targets with power densities (neutrals plus occasionally ions) relevant to DEMO steady state operation, in the range of 20 MW/m2. OLMAT design will allow a series of experiments that other divertor simulator devices cannot easily perform: in-situ measurements of hydrogen retention, redeposition, vapor shielding, material fatigue, dust and precipitates effects, etc. Moreover, a high-power fiber laser will be used to simulate Edge Localized Modes in a small area, or to simulate the strike point power deposition profile.

Xu W., Hu J.S., Sun Z., Maingi R., Zhang L., Yu Y.W., Li C.L., Zuo G.Z., Qian Y.Z., Huang M., Meng X.C., Gao W., Duan Y.M., Chen Y.J., Wang K., et. al.