V.B. Petrov

Energy removal and MHD performance of lithium capillary-pore systems for divertor target application

Abstract 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.

Peculiarity of deuterium ions interaction with tungsten surface in the condition Imitating combination of normal operation with plasma disruption in ITER

Tungsten is a candidate material for the ITER divertor. For the simulation of ITER normal operation conditions in combination with plasma disruptions samples of various types of tungsten were exposed to both steady-state and high power pulsed deuterium plasmas. Tungsten samples were first exposed in a steady-state plasma with an ion current density 10 21 m -2 s -1 up to a dose of 10 25 m - 2 at a temperature of 770 K. The energy of deuterium ions was 150 eV. The additional exposure of the samples to 10 pulses of deuterium plasma was performed in the electrodynamical plasma accelerator with an energy flux 0.45 MJ/m 2 per pulse. Samples of four types of tungsten (W-1%La 2 O 3 , W-131. monocrystalline W(1 1 1) and W-10%Re) were investigated. The least destruction of the surface was observed for W(1 1 1). The concentration of retained deuterium in tungsten decreased from 2.5 x 10 19 m - 2 to 1.07 × 10 19 m -2 (for W(1 1 1)) as a result of the additional pulsed plasma irradiation. Investigation of the tungsten erosion products after the high power pulsed plasma shots was also carried out.

Experimental and calculated basis of the lithium capillary system as divertor material

First results as experimental and calculated basis of a new concept are described in the paper. Experimental models of liquid lithium capillary structure have been tested at long-pulse high heat loads. The power loads on the capillary target up to 50 MW/m2 were provided by an electron beam with electron energy ≤ 9 ke V in a longitudinal magnetic field of 0.25 T. Seven experiments were performed with the different capillary targets. The effects of disruption discharges in tokamaks have been simulated by means of magnetized plasma flows with pulse length of 0.2 ms, electron density of 1022 m3 and energy density up to 4 MJ/m2. The plasma flow was generated by a quasistationary plasma accelerator and interacted with a lithium capillary structure. 2D modelling of the ITER divertor with a lithium target is presented as the first step in the validation of a new divertor concept. A lithium radiative divertor scenario has been examined for the ITER using DDIC95 code. First calculations have shown that thermal loads on the divertor plates are reduced down to 1.3 MW/m2. The main power is radiated in the divertor.

Plasma impact on materials damaged by high-energy ions

We present a short review of experimental research carried out at the NRC Kurchatov Institute over recent years on the behavior of plasma-facing materials (PFMs) when a high level of radiation damage in plasma. Neutron-induced damage was modeled with accelerated ions (in the MeV range) and covered a 1–80 dpa interval. Irradiated carbon materials and tungsten were exposed to deuterium steady-state plasma at deuterium ion energies of 100–250 eV. The work focused on the damaging effect on erosion and on deuterium retention in irradiated materials. The influence of displacement damage was found on the erosion of carbon materials after their bombardment with C+ ions. Changes in deuterium retention were observed on tungsten damaged by 3–4 MeV helium ions. The experiments and results show the efficiency of the method for investigating plasma influence on PFMs for fusion applications taking into account the effect of accumulated radiation damage.

Bulk retention of deuterium in graphites exposed to deuterium plasma at high temperature

Abstract A highly ionized deuterium plasma with a low residual gas pressure and a high intensity D2+-ion beam were used for the study of deuterium retention in RG-Ti-91 and POCO AXF-5Q graphites. Deuterium retention in the samples was estimated by TDS during heating to 2000 K. Mechanical removal of a surface layer 100 μm thick was used to distinguish bulk and surface fractions of retained deuterium. The samples of RG-Ti and POCO graphites were exposed to a plasma with an ion flux of 3 × 1017 D/cm2 · s in the ‘Lenta’ plasma device for 10 to 104 s at residual deuterium pressure of 0.04 Pa at 1400 K. Under plasma exposure deuterium capture in RG-Ti graphite reached the saturation level at a fluence of 4 × 1020 D/cm2 while the bulk inventory was negligible. As for POCO graphite, deuterium retention increased with fluence and was equal to 18 appm in the bulk for a fluence of 7 × 1021 D/cm2. The same amount of deuterium in the bulk was obtained after gas exposure of POCO at an effective pressure of 0.8 Pa (1400 K, 6 h). With this result, the tritium concentration in the plasma-facing graphite materials can reach 1500 appm or 380 grams of tritium per ton of graphite. To understand the role of ion flux in generation of effective pressure, POCO was irradiated with 16 keV D2+-ions at 1400 K for 4 h to 8 × 1020 D/cm2 (ion flux was 6 × 1016 D/cm2 · s, residual deuterium pressure was 0.004 Pa). The results are discussed on the basis of structural differences for POCO and RG-Ti graphites.

Experimental study of lithium target under high power load

Abstract This paper presents experimental research on simulation of a free liquid lithium surface under high heat flux impact for divertor application. Capillary porous structure (CPS) was taken to form the free liquid metal surface imitating divertor target plate. Experiments were performed in the SPRUT-4 linear plasma device with electron beam as power source. Lithium-filled targets were investigated at 1–50 MW / m 2 heat loads in steady state. Lithium evaporation, energy and mass balance, surface temperature, vapor ionization, lithium plasma parameters and radiation were studied. Detailed thermal analysis was made to study heat flows in the target and their correspondence with experimental observations. Durable operation of the setup was possible in the range 1–20 MW / m 2 without damage of the structure. The relevance of the experimental performance to divertor condition is analyzed.

Research of the capillary structure heat removal efficiency under divertor conditions

Abstract Experimental models of capillary structure for liquid metal fusion reactor divertor simulation have been designed, manufactured and tested in order to estimate the behaviour and possibilities of plasma-facing components based on lithium capillary system at long-pulse high heat load. The power load on the capillary target structures up to 50 MW/m 2 was provided by electron beam with electron energy ≤ 10 keV. The exposition-time was up to several minutes and was limited by the lithium quantity in the supply vessel. The operation parameters of the models determined in the experiments are in accordance with there design estimations. The tests of various model constructions at the divertor relevant power loads have shown promise for the new concept of a divertor taking into account long life and reliability.

Research of lithium capillary-pore systems for fusion reactor plasma facing components

Abstract To date there is no adequate solution for high heat load plasma facing components of the next step fusion reactor among solid material options. A lithium-filled capillary porous systems (CPS) was proposed as a plasma facing material and experimental work on this subject is now in progress. Steady-state experiments with CPS-based target and lithium supply systems have shown successful operation at heat fluxes of 1–10 MW/m2 during several hours. Experimental data is obtained on lithium CPS stability at heat flux up to 25–50 MW/m2. The lithium CPS behaviour in contact with real tokamak plasma is considered for normal discharge condition at 10 MW/m2 and for plasma disruption at 15 MJ/m2. Erosion mechanism of lithium under tokamak plasma impact was analysed. Stability of lithium CPS in tokamak conditions was shown.

Liquid lithium surface research and development

Abstract A liquid metal surface made with a capillary porous structure (CPS) (solid base) filled with lithium (liquid) has been offered for application in a magnetic confinement fusion reactor. The approach is investigated experimentally for divertor and first wall relevant conditions. The CPS ensured stability of the liquid surface under pulsed plasma impact in disruption simulation and tokamak experiments. Continuous operation of lithium capillary target was studied under electron beam load in stationary thermal conditions in the range 1–10 MW/m 2 of energy flux in steady state. Lithium evaporation was shown to dominate at temperatures higher than 400 °C and it removed up to 0.7 of incident power. Heat flux redistribution at the liquid lithium surface was analyzed. Lithium ionization, lithium plasma parameters near the liquid surface were evaluated. The importance and possibility of prompt lithium removal from the near surface layer in divertor conditions are emphasized.

Experimental modelling of plasma–graphite surface interaction in ITER

Abstract The investigation of graphite erosion under normal operation ITER regime and disruption was performed by means of exposure of RGT graphite samples in a stationary deuterium plasma to a dose of 10 22 cm −2 and subsequent irradiation by power (250 MW/cm 2 ) pulse deuterium plasma flow imitating disruption. The stationary plasma exposure was carried out in the installation LENTA with the energy of deuterium ions being 200 eV at target temperatures of 770°C and 1150°C. The preliminary exposure in stationary plasma at temperature of physical sputtering does not essentially change the erosion due to a disruption, whereas exposure at the temperature of radiation enhanced sublimation dramatically increases the erosion due to disruption. In the latter case, the depth of erosion due to a disruption is determined by the depth of a layer with decreased strength.