Vapor Shield Research Articles

Effects of Vapor Shield Expansion on Vapor Shield Effectiveness and Plasma Gun Efficiency

The vapor shield outward expansion rate can be shown to affect energy transport through the vapor shield, thereby influencing the vapor shield effectiveness. To more accurately determine the divertor plate erosion depth from a tokamak fusion reactor disruption or plasma gun sources, it is then necessary to include source plasma (beam) momentum transfer and beam mass deposition to the expanding vapor shield. Other factors such as incident heat flux and target Z value are shown to influence the vapor shield expansion rate as well. Code calculations show that increasing heat fluxes can increase the fraction of vapor shield kinetic energy and lower the fraction f of incident energy transported to the solid. Low-Z materials give higher kinetic energies as well but result in a higher f due to a lower specific heat. These results can also be applied to plasma gun technology to help increase its efficiency. In an electrothermal gun, the plasma expansion rate (rate at which vaporized material travels out of the gun) can cause differing plasma residence times and differing plasma temperatures as well. Determining the mechanisms that influence the vapor shield expansion rate and showing its sensitivity on f can give us a qualitative way of determining how changing parameters can influence plasma gun efficiency. Low-energy (<200 eV) disruption plasmas add much mass as well as momentum to a vapor shield. Mass addition can cause the vapor shield temperature and f to differ for a given incident heat flux and change the vapor shield expansion rate as well. Also, we find that deuterium’s shielding effectiveness differs from carbon.

A Model of Turbulent Mixing of Species, Momentum, and Energy in a Vapor Shield Plasma

Evidence suggests that turbulent mixing may affect the energy transport through the vapor shield (VS) formed during a tokamak disruption. The VS is first found to be very unstable according to the Rayleigh-Taylor stability criteria. Adding beam mass to the vaporized material and then mixing of the entire VS is found to cause a significant increase in energy transport through the VS for fusion reactor disruption-relevant energy beams. Mixing the VS for electrothermal gun-relevant energy beams can also affect the energy transport rate, depending on the source species used.

Experimental comparison of carbon and beryllium as divertor target materials in JET

The performance of carbon (CFC) and beryllium as divertor target materials has been investigated in the JET pumped divertor under a wide range of experimental conditions. In general, the characteristics of L- and H-mode plasmas on the two targets were very similar. The impurity content in the two cases was also similar and it is believed that carbon sputtered from plasma-facing components in the main chamber played a significant role in beryllium target experiments. The success of the target design was confirmed by the absence of carbon blooms and beryllium melting on tile edges. However, in a specific melt experiment designed to assess the behaviour of plasmas on molten and melted beryllium, little evidence of vapour shielding of the target was found and melt damage to a depth of 3 mm was observed.

Erosion lifetime of ITER divertor plates

For ITER divertor operation in the detached regime, erosion of divertor plates made of beryllium, carbon fibre composites and tungsten-rhenium alloys is evaluated. Erosion due to (a) physical and, for CFC, chemical sputtering (including surface temperature variation over the lifetime because of neutron irradiation), (b) disruption thermal quench using published data and (c) evaporation and melt layer loss (metals) for infrequent high-power transients (20 MW/m 2, up to 10 s) is evaluated both with and without the reduction of net incident power by radiation from the evaporated impurities (‘low density vapour shield’). The composite lifetime is calculated for normal operation (detached, 10% transients, 10% disruptions); the effect of an increase in transient frequency and in incident power (semi-attached operation) is determined. The resulting erosion lifetime excludes beryllium, whereas both CFC and W-3%Re yield acceptable lifetimes.

Lifetime of Be-, CFC- and W-armoured ITER divertor plates

The thermal performance of divertor plates armoured with beryllium (thickness 2–30 mm), carbon fibre composite (CFC, 2–60 mm) or tungsten (W3Re, 2–60 mm) was modelled for transients of 10 s at 20 MW/m 2 — with and without vapour shielding. Due to the high resulting temperatures special attention was paid to the material properties up to and above melting or sublimation. Pure copper was chosen for the flat heat sink of Be and W3Re, the coolant tube material was DS-Cu in all cases. The melt layer depth, the surface temperature and peak power densities to the coolant were obtained in full 2D geometry using finite element modelling. The code was modified to include melting. Evaporation and vapour shielding at elevated temperatures improve the cooling efficiency and reduce the incident power (converting it to radiation, some of which is radiated elsewhere) and thus reduce the depth of melting. For Be armour 10 mm thick, slow plasma transients at 20 MW/m 2 after 10 s will induce melting to a depth of ∼ 2 mm without vapour shielding, and ∼ 0.15 mm with vapour shielding. For up to 60 mm of CFC the peak temperatures attained stay below sublimation; for up to 20 mm of W3Re the surface temperature stay below melting. The lifetime of the divertor plates is determined from the reduction of thickness due to these transients (10% of shots, 10 s), as well as from disruptions (10% of shots) and sputtering. When the maximum or initial armour thickness is given either by the maximum permitted surface temperature during steady state operation at 5 MW/m 2 (which is 1050 K for Be, permitting 11 mm armour thickness, 1780 K for carbon, permitting 40 mm of CFC after neutron irradiation), or in the case of W3Re by the requirement to stay below melting after 10 s at 20 MW/m 2 (permitting 20 mm), and assuming a minimum remaining thickness of 2 mm for all three armour materials, the predicted lifetimes without vapour shielding during transients are as follows: 120–225 shots for Be, 5840–8170 shots for CFC with moderate chemical sputtering, and 2370–7740 shots for W3Re.

Comprehensive Safety Analysis Code System for Nuclear Fusion Reactors IV: Preliminary Estimation of Dust Production Due to Plasma Disruptions

Dust production due to plasma disruptions has been investigated using a safety analysis code, which can calculate the plasma dynamics and thermal characteristics of fusion reactor structures simultaneously. We selected two fusion reactor designs in the International Thermonuclear Experimental Reactor (ITER ), i.e., the Engineering Design Activity (EDA) and the Conceptual Design Activity (CDA). The ITER/EDA will adopt beryllium for the plasma-facing component (PFC), and the ITER/CDA adopted graphite for PFC. The beryllium dust production in the ITER/EDA reactor will range from 7.0 to 10.3 kg/disruption, which strongly depends on vapor shield effects. The carbon dust production in the ITER/CDA reactor will range from 1.9 to 2.4 kg/disruption. However, the carbon dust will increase by as much as a factor of 2 to 5 because the effective latent heat of graphite has a large uncertainty under the extremely high heat flux during disruptions. For both, dust production from the first wall depends on the current quench time during disruptions. If the current quench time can be extended, the beryllium dust from the first wall will be minimized, and the carbon dust from there will be negligible.

Role of radiation in vapor shielding of first wall during disruption

An initially ablated vapor may provide an important radiation shield to reduce the heat load and the further ablation of the Be divertor strike plate during an ITER disruption. Thus, we investigate the role of line emission in Be at relevant temperatures and densities. For an ion density of 10 18 cm −3 (10 24 m −3), we calculate that there is significant line emission and re-absorption in the temperature range from 3 to 10 eV. The re-absorption limits the line emission that strikes the plate. Line transfer calculations coupled with detailed atomic modeling predict that lines contribute 10–50% of the impinging plate radiation. We note that because of this effect, previous calculations using multi-group radiation transfer may under-estimate the vapor shielding for these ITER-relevant parameters.

Experimental simulation and numerical modeling of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and are experimentally analyzed at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions.

Test of divertor materials under simulated plasma disruption conditions at the SOM electron beam facility

The SOM electron beam facility at the Efremov Institute at St. Petersburg was used for exploratory plasma stream target experiments under conditions aiming the thermal quench phase of ITER tokamak plasma disruptions. The SOM facility uses a vacuum diode with multipoint explosive emission cathode for e-beam formation and a downstream increasing magnetic field for beam transportation and focusing onto the target. The beam energy is around 100 keV. Pulse durations up to 200 μs and energy densities at the target up to 4.5 MJ/m 2 are achieved. Optical interferometry and spectroscopy allowed to study the formation and properties of the plasma formed from vaporized target materials in front of the target. Target erosion was also determined. For carbon erosion starts after reaching a threshold energy density value of 1 MJ/m 2. The onset of target plasma formation strongly depends on beam power density and for 2 MW/cm 2 starts only after 80 μs. Temperature and density of the target plasma remain low and the fraction of neutrals is high. Modeling was done with a radiation hydrodynamic code in one-dimensional planar geometry. Multifrequency radiation transport was treated in forward reverse transport approximation. The modeling results in terms of vapor shield formation, target plasma temperature and erosion are in good agreement with the experimental results. Vapor shield formation is negligible. The erosion is mainly determined by the power density of the unattenuated beam.

Vacuum UV spectroscopy of armor erosion from plasma gun disruption simulation experiments

Abstract Extensive simulations of tokamak disruptions have provided a picture of material erosion that is limited by the transfer of energy from the incident plasma to the armor solid surface through a dense vapor shield. Two transmission grating vacuum ultraviolet (VUV) spectrographs were designed and utilized to study the plasma-material interface in plasma gun simulation experiments. Target materials included POCO graphite, ATJ graphite, boron nitride and plasma-sprayed tungsten. Detailed spectra were recorded with a spatial resolution of ca. 0.7 mm resolution on VIKA at Efremov and on 2MK-200 at Troitsk. Time-resolved data with 40–200 ns resolution were then recorded along with the same spatial resolution on 2MK-200. The VIKA plasma gun directly illuminated a target with a high-intensity plasma pulse of 2–100 MJ m−2 with low-energy ions of ca. 100 eV. The 2MK-200 plasma gun illuminated the target via a magnetic cusp that permitted only deuterium to pass with energies of ca. 1 keV, but which produced a fairly low energy density of 2 MJ m−2. Power densities on target ranged from 107 to 108 W cm−2. Emitted spectra were recorded from 15 to 450 A over a distance from 0 to 7 cm above the armor target surface. The data from both plasma gun facilities demonstrated that the hottest plasma region was sitting several millimeters above the armor tile surface. This apparently constituted the absorption region, which confirmed past computer simulations. Spectra indicated both the species and ionization level that were being ablated from the target, demonstrating impurity content, and showing plasma ablation velocity. Graphite samples clearly showed C V lines as well as impurity lines from O V and O VI. The BN tiles produced textbook examples of B IV and B V, and extensive N IV, V and VI lines. These are being compared with radiation-hydrodynamic calculations.

Review of component erosion in electric launcher technology

Component erosion in electric launcher technology (electromagnetic, electrothermal and electrothermal-chemical) plays a major role, which determines the launcher's bore lifetime, launch velocity and operational limits. Erosion studies have been conducted on the electrothermal launcher SIRENS, which simulates expected operational conditions of high heat flux deposition in electric launchers. Various material surfaces have been exposed to SIRENS plasma to test their performance, erosion behavior and lifetime. Tested materials are pure metals, alloys, refractory materials, specially-coated materials, insulators and composites, pure and composite graphites. Results showed that coating reduced erosion. Tungsten-based alloys have no observable erosion. Vapor shield phenomena have been characterized for different materials, and the energy transmission factor through the vapor shielding layer is obtained. >

Transport of Energetic Ion and Electron Energy Through the Vapor Shield During a Tokamak Plasma Disruption

Energetic (> 10-keV) particles incident on divertor plate surfaces may penetrate the vapor shield formed under extremely high heat flux conditions (> 1010 W/m2). In this case, the total energy tran...

Upgrade of the Vapor Shield Modeling Code MAGFIRE

An upgrade of the MAGFIRE code used to model the vapor shield (VS) effect was needed to include: higher disruption heat fluxes of interest to ITER, a comparison of carbon to beryllium erosion depths and the effects of ion/electron beam charge separation in the VS. Results show that higher heat fluxes give a lower total energy transmission factor (f) as expected. Beryllium divertor plates have much higher erosion depths than do carbon as well as a higher f. Charge separation has a small effect on the VS stopping power for electrons and on the distribution of deposited energy in the VS. However, the effect will be more important as the disruption particle energy increases.

Numerical modeling and experimental simulation of vapor shield formation and divertor material erosion for ITER typical plasma disruptions

The high divertor heat load during a tokamak plasma disruption results in sudden evaporation of a thin layer of divertor plate material, which acts as vapor shield and protects the target from further excessive evaporation. Formation and effectiveness of the vapor shield are theoretically modeled and experimentally investigated at the 2MK-200 facility under conditions simulating the thermal quench phase of ITER tokamak plasma disruptions. In the optical wavelength range C II, C III, C IV emission lines for graphite, Cu I, Cu II lines for copper and continuum radiation for tungsten samples are observed in the target plasma. The plasma expands along the magnetic field lines with velocities of (4 ± 1) × 10 6 cm/s for graphite and 10 5 cm/s for copper. Modeling was done with a radiation hydrodynamics code in one-dimensional planar geometry. The multifrequency radiation transport is treated in flux limited diffusion and in forward reverse transport approximation. In these first modeling studies the overall shielding efficiency for carbon and tungsten defined as ratio of the incident energy and the vaporization energy for power densities of 10 MW/cm 2 exceeds a factor of 30. The vapor shield is established within 2 μs, the power fraction to the target after 10 μs is below 3% and reaches in the stationary state after about 20 μs a value of around 1.5%.

Plasma gun experiments and modeling of disruptions

Potentially high erosion due to ablation from plasma disruptions looms as a nemesis for advanced fusion devices such as ITER. Under some conditions, it is believed that the material ablated during a disruption forms a “vapor shield” that mitigates subsequent ablation. This paper presents new experimental data that identifies an absorption surface in the ablating plasma that is above the irradiated armor surface. We also present a summary of progress in modeling using a 1-D Lagrangian hydrodynamics code. Experimental erosion data will be reviewed from the PLADIS facility, a plasma gun at the University of New Mexico on several materials, including Be, tungsten, copper, graphites. Profile measurements of the crater topology showed the eroded Be surface to be much rougher than that of carbon and to demonstrate erosion rates that were almost factors of four greater than graphite. Plasma guns at the D.V. Efremov Scientific Research Institute and at TRINITI, both in the Russian Federation, are being utilized to confirm spectroscopic vapor shield data.

The use of an electrothermal plasma gun to simulate the extremely high heat flux conditions of a tokamak disruption

Disruption damage conditions for future large tokamaks like ITER are nearly impossible to simulate on current tokamaks. The electrothermal plasma source SIRENS has been designed, constructed, and operated to produce high density (> 1025/m3), low temperature (1–3 eV) plasma formed by the ablation of the insulator with currents of up to 100 kA (100 μs pulse length) and energies up to 15 kJ. The source heat fluence (variable from 0.2 to 7 MJ/m2) is adequate for simulation of the thermal quench phase of plasma disruption in future fusion tokamaks. Different materials have been exposed to the high heat flux in SIRENS, where comparative erosion behavior was obtained. Vapor shield phenomena has been characterized for different materials, and the energy transmission factor through the shielding layer is obtained. The device is also equipped with a magnet capable of producing a parallel magnetic field (up to 16 T) over a 8 msec pulse length. The magnetic field is produced to decrease the turbulent energy transport through the vapor shield, which provides further reduction of surface erosion (magnetic vapor shield effect).

Magnetic vapor shielding mechanism in electromagnetic and electrothermal launchers

The vapor shield formed by the ablation of wall materials in plasma-driven launchers absorbs a large fraction of the incoming energy and reduces the surface erosion. Decreasing the turbulent energy transport through the vapor shield may result in further reduction of the surface erosion. Applying a strong magnetic field parallel to the surface (magnetic vapor shield) is a possible approach. SIRENS, an electrothermal launcher, is equipped with a pulsed high intensity magnet, and has been used to investigate the magnetic vapor shield effect. The magnetic field has been applied at different field intensities (2.6 to 8.75 T), for launcher input energy varying from 1 to 5 kJ. A 25-35% reduction of the surface ablation has been achieved at 6.25 T for energy inputs of 1-2 kJ, with a trend of further reduction at higher values of the magnetic field.< <ETX xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">&gt;</ETX>

Comparative study of component erosion for electromagnetic and electrothermal launchers

The electrothermal launcher SIRENS has been used to study the erosion of critical components (rails and insulators) of plasma-driven launchers. SIRENS can produce high-density (>10/sup 25//m/sup 3/) low-temperature (1-3 eV) plasma, formed by the ablation of the insulator (Lexan), with currents up to 100 kA. The incident heat flux varies between 2 and 90 GW/m/sup 2/ over 100 mu s duration, for input energies of 1-10 kJ. Erosion studies have been performed on several insulators, pure and coated metals, alloys, and several graphite grades. The fraction of the total incident energy transmitted to the eroded surface varies from 12 to 30% for the materials tested and decreases to 5-7% as the incident energy fluence increases. Such reduction in erosion for a given incident fluence is due to the vapor shield effect. The scaling law for the energy transmission factor through the vapor shield layer was obtained for the exposed materials. >

Vapor shield protection of plasma facing components under incident high heat flux

Disruption damage to plasma facing components has been found to be a limiting design constraint in ITER and other large fusion devices. A growing data base is confirming the role of the vapor shield in protecting ablated surfaces under disruption-like conditions, which would imply longer lifetimes for plasma facing components. We present new results for exposure of various material surfaces to high heat fluxes up to 70 GW/m 2 over 100 μs (7 MJ/m 2 ) in the SIRENS high heat flux test facility. Tested materials are graphite grades, pyrolytic graphite, refractory metals and alloys, refractory coatings on copper substrates, boron nitride and preliminary results of diamond coating on silicon substrates. An empirical scaling law of the energy transmission factor through the vapor shield has been obtained. The application of a strong external magnetic field, to reduce turbulent energy transport in the vapor shield boundary, is shown to decrease f by as much as 35% for fields of 8 T.

Numerical analysis of the plasma-material interaction during the thermal quench phase of tokamak disruptions

Plasma disruptions in a next generation tokamak will cause high surface energy deposition on the surface of the plasma facing components. After the onset of the energy deposition a series of complex processes develops which are commonly termed as “vapour shielding”. The vapour shield developing in front of the heated material surface may attenuate a part of the incident energy and cause a reduction in the energy which is actually deposited in the plasma facing material. The present paper describes a numerical model for these processes which includes the thermophysical response of the plasma facing material, the magnetohydrodynamic development of the plasma vapour cloud, and the radiation transfer within the vapour cloud. Results of the calculations for an incident energy of 10 MJ m −2 during 0.1 ms using the local thermodynamic equilibrium model are presented. Compared to a calculation without vapour shield the carbon erosion is reduced by 98.9%.