Surface Heat Load Research Articles

Tungsten Melt Losses under QSPA Kh-50 Plasma Exposures Simulating ITER ELMs and Disruptions

Experimental simulations of ITER transient events with surface heat load parameters relevant to edge-localized-mode (ELM) impacts and disruptions have been performed with a quasi-stationary plasma accelerator Kh-50. In the ELM simulation experiments with heat loads exceeding the tungsten melting threshold, both droplet splashing and solid dust ejection are observed. The erosion products emitted from the exposed tungsten surfaces in the form of droplets and solid dust have been clearly distinguished by variation of impacting heat load with performed analysis of particle ejection start time, their velocities, and changes in the luminosity of the particle traces in front of the target surface recorded with a charge-coupled device. Droplets are emitted during plasma exposure, and dust generation dominates after the end of the plasma pulse, at the time of the following material cooling. The contributions of Kelvin-Helmholtz instabilities to droplet splashing from the melt layer are discussed. Decrease of droplet velocity with increasing surface heat load is observed. This decrease could be attributed to the growing size of the droplets for higher energy loads.

CFD analysis of the ITER first wall 06 panel. Part II: Thermal-hydraulics

The computational fluid dynamics (CFD) analysis of the FW06 panel of the ITER shielding blanket is presented in two companion papers. In this Part II we concentrate on the thermal-hydraulics of the water coolant, driven by the nuclear volumetric and plasma surface heat loads discussed in Part I. Both the detailed steady state analysis of a single cooling channel and the coarse transient analysis of the whole panel are considered. The compatibility of the hot spots with the maximum recommended temperatures for the different materials is confirmed. The heat transfer coefficient between coolant and walls is obtained post-processing the results of the simulation and compared with the results of available correlations, which may be used for simpler analyses: in the fully developed flow regions of the cooling pipes, it turns out to be well approximated by the Sieder–Tate correlation. The operation margin with respect to the critical heat flux is also computed and turns out to be sufficiently large compared with the design limit.

Study on Calculation Method for Partition of Heat Transfer in an Ultra-supercritical Boiler

Use a 1000MW ultra-supercritical tower boiler as the research object. On the basis of one dimensional model, simplify the tube heat transfer model and the radiation heat transfer model; establish the two-dimensional area calculation model with the regional method; summarize the heat load distribution of flue gas temperature and water wall surface; and compare with the measured data. The error range of the result is acceptable on the project. The distribution of water wall surface heat load along the furnace width and the area where heat transfer deterioration cause easily along the furnace height direction are studied with the model and algorithm on different boiler load conditions. All these provide the reference for the design and operation of the ultra supercritical boiler.

Hypersonic environment assessment of the CIRA FTB-X re-entry vehicle

This paper deals with design activities performed by CIRA in the frame of its flying test bed vehicle development, named FTB-X. The re-entry scenario with the corresponding loading environment for the proposed vehicle, devoted to research activities in re-entry technologies, is reported and analyzed. The hypersonic environment characteristics of FTB-X are investigated on a trajectory-based design approach by means of several engineering and CFD analyses. Being a phase A design, the air is modeled only as a perfect gas and the analyses were based on laminar flow conditions, while only a preliminary analysis of real gas and turbulence effects has been performed in time. The results show that the aerodynamics derived from engineering design approach is sufficiently accurate for preliminary analysis purpose; moreover, surface heat loads were computed on the vehicle configuration for thermal shield design scopes.

Russian development of enhanced heat flux technologies for ITER first wall

Recently the ITER first wall (FW) design has been significantly upgraded to improve resistance to electromagnetic loads, to facilitate FW panel replacement and to increase FW ability to withstand higher (up to 5MW/m2) surface heat loads. The latter has made it necessary to re-employ technologies previously developed for the now-abandoned port limiters. These solutions are related to the cooling channel with CuCrZr–SS bimetallic walls and hypervapotron type cooling regime, optimization of Be-tiles dimensions and Be to CuCrZr joining technique. A number of representative mockups were tested at high heat flux (HHF) at the Tsefey electron-beam facility to verify the thermo-hydraulic characteristics of the reference cooling channel design at moderate water flow velocities (V=1–3m/s, P=2–3MPa, T=110–170°C). The heat flux was gradually varied in the range of 1–10MW/m2 until the critical heat flux was registered. The mockups of hypervapotron structure demonstrated the required cooling efficiency and critical heat flux margin (1.4) at a water velocity of ≥2m/s. Dimensions of Be armor tiles strongly affect the thermo-mechanical stresses both in the CuCrZr cooling wall and at the Be–CuCrZr interface. Results of tile dimensions optimization (variable in the range 12mm×12mm×6 to 50mm×50mm×8mm) obtained by the HHF (variable in the range of 3–8MW/m2) experiments are presented and compared with analysis. It is shown that optimization of the tile geometry and joining technology provides the required cyclic fatigue lifetime of the reference FW design.

Estimation of the dust production rate from the tungsten armour after repetitive ELM-like heat loads

Experimental simulations for the erosion rate of tungsten targets under ITER edge-localized mode (ELM)-like surface heat loads of 0.75 MJ m−2 causing surface melting and of 0.45 MJ m−2 without melting have been performed in the QSPA-Kh50 plasma accelerator. Analytical considerations allow us to conclude that for both energy deposition values the erosion mechanism is solid dust ejection during surface cracking under the action of thermo-stress. Tungsten influx into the ITER containment of NW∼5×1018 W per medium size ELM of 0.75 MJ m−2 and 0.25 ms time duration has been estimated. The radiation cooling power of Prad=150–300 MW due to such influx of tungsten is intolerable: it should cool the ITER core to 1 keV within a few seconds.

Simulation of ITER edge-localized modes' impacts on the divertor surfaces within plasma accelerators

Experimental simulations of ITER transient events with relevant surface heat load parameters (energy density and pulse duration) as well as particle loads were performed with a quasi-steady-state plasma accelerator Kh-50 and pulsed plasma guns. Additional shielding was observed during irradiation of the combined carbon–tungsten (C–W) surface. An evaporated C cloud protects W from evaporation even at an energy density of 2.4 MJ m−2 in impacting plasma. The spectrum of tungsten and parameters of shielding layers were found under powerful exposure of the tungsten surface. An influence of material modification through plasma exposures on cracking thresholds of tungsten is emphasized. It was found that increasing the number of exposures to more than 20 plasma pulses of 0.25 ms in duration shifts the energy threshold for crack development from 0.3 down to 0.2 MJ m−2. Differences in the evolution of tungsten substructure after exposure to helium and hydrogen plasma streams of different duration are also analyzed.

Spectroscopy of Plasma Surface Interaction in Experiments Simulating ITER Transient Events

This report presents the results of experimental studies of powerful plasma impacts upon tungsten and carbon surfaces, which are ITER relevant Plasma Facing Materials (PFMs). The simulation experiments were carried out with a QSPA Kh-50 (Quasi-Stationary Plasma Accelerator) in Kharkov Institute of Physics and Technology (Ukraine) and an RPI-IBIS (Multi-Rod Plasma Injector) facility in the Andrzej Soltan Institute for Nuclear Studies (IPJ) Swierk (Poland).QSPA Kh-50 generates hydrogen plasma streams of duration of 0.25 ms and the heat loads in the range of 0.2–2.5 MJ/m2, which correspond to the Edge Localized Modes (ELM) impacts expected in ITER. The plasma stream diameter is 18 cm, averaged ion energy is about 0.4 keV, and the maximum plasma pressure achieves 3.2 bar. Due to that fact, using the QSPA Kh-50 it is possible to simulate ITER transient events. Deuterium plasma streams with power density of 10-50 W/m2 and pulse duration of 1-5 μs, generated by RPI-IBIS were used for comparative studies and determination of an initial stage of evaporated impurities dynamics during plasma-surface interactions as well as features of surface damages appearing under varied plasma parameters.In order to determinate the main plasma parameters (an electron density and temperature) and to study of impurities behavior at the time of discharge the use was made of optical spectroscopy methods. The onset of a vapor shield in front of the target surface was investigated in dependence on a surface heat load for tungsten (W) and carbon (C) targets. Information about dynamics of the W- and C-ions production was obtained.Some issues of the droplet splashing at the tungsten surfaces and the formation of hot spots upon the graphite surface, which can be sources of the enhanced evaporation, are also discussed.

Thermomechanical Analysis of the Revised U.S. ITER DCLL Test Blanket Module

The US Fusion Nuclear Science and Technology program selected the Dual Coolant Lead Lithium (DCLL) concept as the primary Test Blanket Module (TBM) for testing in ITER. The DCLL blanket concept has the potential to be a high-performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is the structural material, helium is used to cool the first wall and blanket structure, and the self-cooled Pb-17Li breeder is circulated for power conversion and tritium extraction.The DCLL TBM has undergone major design changes since 2005. We present here the most recent thermo-mechanical analysis of the newly revised DCLL TBM. The analysis described here is aiming to verify the thermo-mechanical response of the DCLL TBM under relevant normal operating conditions as well as during a loss of coolant accident (LOCA).A full 3-dimensional solid model of the entire DCLL TBM structure was developed, which included FW, top and bottom lids, internal supporting ribs, manifolds, plena, and flexible frame-attachment supports. A coupled thermo-mechanical analysis was performed for both normal- and off-normal operating conditions. Thermal loads included surface heat load, volumetric heating, as well as detailed position- and location dependent heat transfer along all coolant channels. Structural loads incorporated helium coolant pressure loads, self-weight, as well as the weight of the PbLi. Maximum structure temperatures of nearly 560 °C along with a maximum resultant net displacement of more than 10 mm were mapped for normal operating conditions and a number of stress concentration locations were identified. The ITER SDC-IC-1300 criteria were applied to the LOCA analysis results. It is shown that the DCLL TBM exhibits admissible behavior regarding the ITER Design Criteria and that the most recent design modifications did not compromise the structural integrity.

A subcooled boiling heat transfer predictive model for ITER EHF FW designs

Recent experimental data from the ITER critical heat flux (CHF) mock-ups was used to benchmark a 3D CFD code concerning subcooled boiling heat transfer for high heat flux removal. The predicted temperatures show good agreement with experimental measurements for a range of operating parameters and of cooling configurations. Specifically, it applies to a hypervapotron channel exposed to a 5 MW/m 2 surface heat load and cooled by velocity of 2 m/s. Such flow geometry and operating condition seem necessary for ITER-enhanced heat flux first wall modules if an adequate design margin in CHF is needed. A detailed CFD and heat transfer analysis performed on a prototyped CAD model provided a higher confidence on the design and is deemed a desirable feature for continued design exploration and optimization processes. This is particularly crucial in regard to flow distribution among the FW fingers.

On temperature bifurcation of beryllium and lithium plasma facing components

The mechanism of temperature bifurcation of plasma contacting surfaces due to recycling of the ionized surface material vapor is considered. It is shown that this mechanism can lead to overheating of beryllium and lithium plasma facing components (in particular, in fusion devices) prior to the thermionic electron emission mechanism. The surface temperatures and the plasma parameters, at which the considered mechanism triggers the local overheating of beryllium and lithium components, are evaluated. The increase in the surface heat load due to secondary electron emission is also considered. It is shown that the combined effects of energy and impact angle distributions of the plasma electrons can increase the averaged secondary electron emission yield to values higher than unity and can lower the average electron energy, at which such yields are achieved.

Experimental and Numerical Investigation of the Thermal Performance of the Gas-Cooled Divertor Plate Concept

The helium-cooled plate-type divertor concept proposed by Malang was designed to accommodate a surface heat load of ˜10 MW/m2. This design can potentially reduce the number of modules needed for the divertor by over two orders of magnitude compared with other concepts, thereby significantly reducing coolant delivery system complexity and manufacturing costs. While previous analyses have predicted that the plate design can accommodate heat fluxes of 10 MW/m2, no experimental data have been published to date to validate such analyses. Experiments have therefore been conducted using air as the coolant at Reynolds numbers similar to those proposed for the actual helium-coolant operating conditions on an instrumented test module with cross-sectional geometry identical to the prototypical plate-type divertor. A second test module where the planar jet exiting the inlet manifold is replaced by a two-dimensional hexagonal array of circular jets over the entire top surface of the inlet manifold has also been tested. The thermal performance of both test modules with and without a porous metallic foam layer in the gap between the outer surface of the inlet manifold and the cooled surfaces was directly compared to test the numerical simulations of Sharafat which predict that the metallic foam significantly enhances heat transfer.

Effect of micrometric hot spots on surface temperature measurement and flux calculation in the middle and long infrared

The non-uniformity of the target temperature due to micrometric hot spots (Hermann et al 2004 Phys. Scr. T 111 98) is an explanation for the experimental fact that near-infrared measurements yield higher temperature values than mid-infrared measurements (Hildebrandt et al 2003 InfraMation 2003 Proc. (Las Vegas, USA, October 2003), Delchambre et al 2005 J. Nucl. Mater. 337–339 1069). The issue of micrometric hot spot disturbance in the surface temperature (Tsurf) measurement and heat load calculation is addressed in this paper. The theoretical investigation at 3, 5 and 12 µm and experiments in the range 3.5–5 µm indicate that the surface state can play an important role in the non-uniform heating surface and consequently in the overestimation of the bulk temperature. The contribution of the hot spots to temperature measurements and flux calculations has been simulated at different wavelengths. Calculations show that (1) the overestimation of the bulk temperature decreases with the wavelength and (2) the overestimation depends on the temperature difference, ΔT, between the bulk and the micrometric hot spots. In addition, experiments have been carried out in order to compare the flux calculations at different wavelengths on different graphite (polished, dusty). The results obtained are very sensitive to the surface state pointing out the difficulties in improving the heat flux calculation model, since the surface state can change during the plasma discharges. This paper shows that the problem of non-homogenous surface temperature can be significantly diminished on working at longer wavelengths.

Experimental study of plasma energy transfer and material erosion under ELM-like heat loads

Main features of plasma–surface interaction and energy transfer to tokamak plasma facing components are studied at different heat loads in ELM simulation experiments with the plasma gun QSPA Kh-50. Repetitive plasma exposures of tungsten, graphite and different combined W-C targets were performed at the pulse duration of 0.25 ms and the heat loads varied in the range 0.2–2.5 MJ/m 2. The onset of vapor shield in front of the surface was investigated. The evaporation is immediately followed by a saturation of surface heat load if further increasing the impact energy. The presence of graphite essentially decreases the heat flux to the nearby tungsten surface, which is due to the carbon vapor shield. Droplet splashing at the tungsten surface and formation of hot spots on the graphite surface are discussed.

Determination of a thermal property of carbon layers from IR measurements in JET NBI test bed using optimisation methods

An optimisation algorithm is used to realize an automated determination of a carbon layer thermal property from IR temperature measurements. This algorithm is based on a 1D linear thermal model of the plasma facing component. It is tested on measurements from two MkIIa JET divertor tiles placed in the NBI JET test bed. Heat fluxes of 5–100 MW/m 2 were applied on those tiles. The resulting surface distribution of the carbon layer is discussed, including its evolution through the experiment, and its influence on the computed surface heat loads. Heat fluxes computed from IR and thermocouple temperature measurements are compared to the previous results. First computation of heat loads during short high flux NBI pulses simulating ELMs using previously determined carbon layer properties are also presented.

Damage to preheated tungsten targets after multiple plasma impacts simulating ITER ELMs

The behavior of a preheated at 650 °C tungsten targets under repetitive ELM-like plasma pulses is studied in simulation experiments with the quasi-stationary plasma accelerator QSPA Kh-50. The targets have been exposed up to 350 pulses of the duration 0.25 ms and the surface heat loads either 0.45 MJ/m 2 or 0.75 MJ/m 2, which is below and above the melting threshold, respectively. The development of surface morphology of the exposed targets as well as cracking and swelling at the surface is discussed. First comparisons of obtained experimental results with corresponding numerical simulations of the code PEGASUS-3D are presented.

Simulation of dust production in ITER transient events

Dust production from the divertor armour during edge-localised modes (ELMs) has been investigated. Analysis of available experimental data on the dust particle production and the particles distribution on size for the MPG-8 graphite and for NB31 carbon fibre composite (CFC) under the disruption-like surface heat load allowed revealing the unknown mechanical parameters of the NB31 CFC. Using these data the code PEGASUS-3D has been fitted and verified for simulation of the dust production by ELMs. First simulation of the dust production for the ELM of 1 MJ/m 2 heat load and 0.5 ms time duration has been calculated.

Minimal irreversibility and optimal distributions of heat transfer surface area and heat load in heat transfer systems

A minimum possible rate of entropy production together with corresponding distributions of heat transfer surface areas and contact temperatures is estimated for heat transfer systems with given values of heat loads and heat transfer coefficients. It is proved that the ratio of the temperatures of contacting flows at any point of the system should be the same when the heat flux in the system is proportional to the difference of temperatures. It is also proved that in this case the temperatures of heating flows at the system outlet should be the same.

Surface temperature measurement and heat load estimation for carbon targets with plasma contact and machine protection

Surface temperature control of actively cooled targets is part of the machine protection in long pulse experiments. The protection of the target is as good as the surface measurement and the thermal model of the target. This paper deduces thermal parameters necessary to correct the measured temperature for surface effects. The temperature measurement error due to the intrinsic CFC structure is discussed. Layers on top of the bulk material result in a change of the temperature response and in an overestimation of the calculated heat flux. All discussed effects did not reduce the machine safety but can reduce the operational space of the experiment. The thermal surface properties have to be checked in situ by changing the heat load to the target and calculating the temperature response.

Shielding design of the ITER NBI duct for nuclear and bremsstrahlung radiation

Three-dimensional Monte Carlo shielding analyses are conducted on the ITER Neutral Beam Injection (NBI) duct for the nuclear and bremsstrahlung radiation. The detailed distribution is evaluated about the nuclear heating rate and surface heat load of the NBI duct wall by the neutron and photon transport calculation. The analytical representations of these nuclear responses are established as a function of the distance from the blanket surface. It is clarified that these representations are different between the duct wall facing the plasma and that hidden from the plasma, and also between the duct wall in the blanket region and that in the vacuum vessel region. These results are very useful for the shielding design of the NBI duct wall in the nuclear fusion reactor.