Characteristics of surface oscillation on high speed liquid Li jet

In the International Fusion Materials Irradiation Facility (IFMIF), high speed liquid lithium (Li) wall jet will be used as target irradiated by two deuteron beams of 125 mA at 40 MeV. To obtain knowledge of Li flow behavior, we have been studying on the surface wave characteristics experimentally using the liquid metal Li circulation loop at Osaka University. In this present study, the characteristic of surface oscillation on high speed liquid Li jet were examined. The free surface oscillation of Li flow was measured by an electro-contact probe apparatus, which detects electric contacts between a probe tip and Li surface. It was installed at 175 mm and 15 mm downstream from the nozzle exit to see influence of the initial growth of surface waves. The wave height of free surface waves was obtained from contact signal. While at 15 mm region the flow surface is very smooth covered with small waves in amplitude, the surface waves are developed sufficiently at the 175 mm. In the case of the velocity of 15 m/s, the maximum wave height reaches 4.8 mm. Heat deposition was estimated on the target back-plate with using the present statistical wave data.     

CFD simulation of the IAEA 10 MW generic MTR reactor under loss of flow transient

Three-dimensional simulation of the IAEA 10 MW generic reactor under loss of flow transient is introduced using the CFD code, Fluent. The IAEA reactor calculation is a safety-related benchmark problem for an idealized material testing reactor (MTR) pool type specified in order to compare calculational methods used in various research centers. The flow transients considered include fast loss of flow accidents (FLOFA) and slow loss of flow accidents (SLOFA) modeled with exponential flow decay and time constants of 1 and 25 s, respectively. The transients were initiated from a power of 12 MW with a flow trip point at 85% nominal flow and a 200 ms time delay. The simulation shows comparable results as those published by other research groups. However, interesting 3D patterns are shown that are usually lost based on the one-dimensional simulations that other research groups have introduced. In addition, information about the maximum clad surface temperature, the maximum fuel element temperature as well as the location of hot spots in fuel channel is also reported.     

Bubble rise characteristics after the departure from a nucleation site in vertical upflow boiling of subcooled water

Rise characteristics of vapor bubbles after the departure from a nucleation site in forced convective subcooled flow boiling were studied visually using two synchronized high speed video cameras. The test section was a transparent glass tube of 20 mm in inside diameter, filtrated and deionized tap water was used as a working fluid, and the flow direction adopted was vertical upward. The outer surface of test section tube was electrically heated to generate vapor bubbles inside of the tube. In the present experiments, the mass flux and liquid subcooling were varied within 94–1435 kg/m² s and 2.2–10 K, respectively. Since the observations were performed at low heat fluxes to avoid the significant increase in the number of active nucleation sites, the obtained bubble images were clear enough to carry out the detailed image analysis for the rise characteristics of individual bubbles. The following three different bubble rise paths were observed after the departure from nucleation sites: some bubbles slid upward the vertical wall for long distance, while other bubbles were detached from the wall after sliding for several millimeters and then migrated toward the bulk liquid; after the migration, some of the detached bubbles were collapsed in subcooled liquid but others remained close to the wall and were reattached to the wall. The results of detailed image analyses suggested that the variation in bubble shape from flattened to more rounded was of primary importance for the occurrence of bubble detachment from the wall.     

Influence of temperature and plasma composition on deuterium retention in refractory metals

Refractory materials are being considered potential candidates to build the first wall of the fusion reactor chamber. This work reports on the results of the study of tungsten and molybdenum metals exposed to high flux densities (～10²⁴ D/m² s) and low temperature (T_e ～ 3 eV) deuterium plasmas in Pilot-PSI irradiation facility.The hydrogenic retention in poly-crystalline W and Mo targets was studied with ³He nuclear reaction analyses (NRA). The NRA results clearly show a two-dimensional radial distribution of the deuterium with a minimum at the center and a maximum close to the edge. These distribution correlates well with the thermal profile of the sample surface, where a maximum of ～1600 K was measured at the center decreasing to ～1000 K in the edges. A maximum deuterium fluence retention of 5 × 10¹⁵ D/cm² was measured. The values of the retained fractions ranging from 10^?5 to 10^?6 D_retained/D_incident were measured with thermal desorption spectroscopy (TDS) and compares well with IBA results. Moreover, the presence of C in the plasma and its co-deposition increases the D retention in the region where a C film is formed. Both NRA and TDS results show no clear dependence of retention on incident fluence suggesting the absence of plasma related traps in W under these conditions.     

Influence of thermal performance on design parameters of a He/LiPb dual coolant DEMO concept blanket design

Spanish Breeding Blanket Technology Programme TECNO_FUS is exploring the technological capabilities of a Dual-Coolant He/Pb15.7Li breeding blanket for DEMO and studying new breeding blanket design specifications. The progress of the channel conceptual design is being conducted in parallel with the extension of MHD computational capabilities of CFD tools and the underlying physics of MHD models. A qualification of MHD effects under present blanket design specifications and some approaches to their modelling were proposed by the authors in [1]. The analysis was accomplished with the 2D transient algorithm from Sommeria and Moreau [2] and implemented in the OpenFOAM CFD toolbox [3]. The thermal coupling was implemented by means of the Boussinesq hypothesis. Previous analyses showed the need of improvement of FCI thickness and thermal properties in order to obtain a desirable liquid metal temperature gain of 300 °C. In the present study, an assessment through sensitivity and parametric analyses of the required FCI thickness is performed.Numerical simulations have been carried out considering a Robin-type thermal boundary condition which assumes 1D steady state thermal balance across the solid FCI and Eurofer layers. Such boundary condition has been validated with a fluid–solid coupled domain analysis.Results for the studied flow conditions and channel dimensions show that, in order to obtain a liquid metal temperature gain of about 300 °C, the required FCI material should have a very small effective heat transfer coefficient ((k/δ) ≤ 1 W/m²K) and fluid velocities should be about 0.2 m/s or less. Moreover, special attention has to be placed on the temperature difference across the FCI layer. However, for a maximised liquid metal thermal gain, higher velocities would be preferable, what would also imply a reduced temperature difference across the FCI layer.     

Simulation of bubbly flow in vertical pipes by coupling Lagrangian and Eulerian models with 3D random walks models: Validation with experimental data using multi-sensor conductivity probes and Laser Doppler Anemometry

A set of two phase flow experiments for different conditions ranging from bubbly flow to cap/slug flow have been performed under isothermal concurrent upward air–water flow conditions in a vertical column of 3 m height. Special attention in these experiments was devoted to the transition from bubbly to cap/slug flow. The interfacial velocity of the bubbles and the void fraction distribution was obtained using 2 and 4 sensors conductivity probes.Numerical simulations of these experiments for bubbly flow conditions were performed by coupling a Lagrangian code with an Eulerian one. The first one tracks the 3D motion of the individual bubbles in cylindrical coordinates (r, ?, z) inside the fluid field under the action of the following forces: buoyancy, drag, lift, wall lubrication. Also we have incorporated a 3D stochastic differential equation model to account for the random motion of the individual bubbles in the turbulent velocity field of the carrier liquid. Also we have considered the deformations undergone by the bubbles when they touch the walls of the pipe and are compressed until they rebound.The velocity and turbulence fields of the liquid phase were computed by solving the time dependent conservation equations in its Reynolds Averaged Transport Equation form (RANS). The turbulent kinetic energy k, and the dissipation rate ? transport equations were simultaneously solved using the k, epsilon model in a (r, z) grid by the finite volume method and the SIMPLER algorithm. Both Lagrangian and Eulerian calculations were performed in parallel and an iterative self-consistent method was developed. The turbulence induced by the bubbles is an important issue considered in this paper, in order to obtain good predictions of the void fraction distribution and the interfacial velocity at different gas and liquid flow conditions.     

Gravitational flow of a thin film of liquid metal in a strong magnetic field

The influence of a poloidal magnetic field of the spherical Tokamak on super thin (h ≈ 0.1 mm) film flow of liquid metal driven by gravity over the surface of the cooled divertor plate is addressed. The experimental setup developed at the Institute of Physics, University of Latvia (IPUL) is described, which makes it possible to drive and visualize such liquid metal flows in the solenoid of the superconducting magnet “Magdalena”. As applied to the above setup, the magnetic field effect on the operation of the capillary system of liquid metal flow distribution (CSFD) is evaluated by using molten metal (lithium or eutectic InGaSn alloy) with a very small linear flowrate q ≤ 1 mm²/s, spread uniformly across the substrate. The magnetic field effect on the main parameters of the fully developed film flow is estimated for the above-mentioned liquid metals.An approximation technique has been proposed to calculate the development of the gravitational film flow. A non-linear differential second order equation has been derived, which describes the variation of the film flow thickness over the substrate length versus the flowrate q, magnetic field B and the substrate sloping α.Results of InGaSn film flow observations in a strong (B = 4 T) poloidal magnetic field are presented. Analysis of the video records evidences of experimental realization of a stable stationary film flow at width-uniform supply of InGaSn.     

Kinetic Monte Carlo simulation of hydrogen diffusion on tungsten reconstructed (0 0 1) surface

A kinetic Monte Carlo (KMC) algorithm has been developed to study the hydrogen diffusion on tungsten reconstructed (0 0 1) surface in the temperature range 220–300 K. The hydrogen diffusion coefficients predicted by the developed KMC code match the experimental values very well at low hydrogen coverage of a fraction of monolayer. A diffusion coefficient formula as a function of temperature and hydrogen coverage was derived from KMC simulations. Due to the very low probability of hydrogen occupying the long bridge adsorption sites, the rates of hydrogen atom having 3 or 4 neighbors are found to be zero for hydrogen coverage much less than a monolayer, while the rates of hydrogen atom having 0–2 neighbors are linear with respect to the hydrogen coverage. The calculated average rates of hydrogen located at the LB sites are very close to zero for low hydrogen coverage. Hydrogen only starts to occupy the LB sites after almost all SB sites are occupied.     

Experimental study and analysis of the purge gas pressure drop across the pebble beds for the fusion HCPB blanket

The lithium ceramic and beryllium pebble beds of the breeder units (BU), in the fusion breeding blanket, are purged by helium to extract the bred tritium. Therefore, the objective of this study is to support the design of the BU purge gas system by studying the effect of pebbles diameter, packing factor, pebble bed length, and flow inlet pressure on the purge gas pressure drop. The pebble bed was formed by packing glass pebbles in a rectangular container (56 mm × 206 mm × 396 mm) and was integrated into a gas loop to be purged by helium at BU-relevant pressures (1.1–3.8 bar). To determine the pressure drop across the pebble bed, the static pressure was measured at four locations along the pebble bed as well as at the inlet and outlet locations. The results show: (i) the pressure drop significantly increases with decreasing the pebbles diameter and slightly increases with increasing the packing factor, (ii) for a constant inlet flow velocity, the pressure drop is directly proportional to the pebble bed length and inlet pressure, and (iii) predictions of Ergun's equation agree well with the experimental values of the pressure drop.     

Experimental investigation of local two-phase flow parameters of a subcooled boiling flow in an annulus

As a series of subcooling boiling flow tests, local two-phase flow parameters were obtained at SUBO (subcooled boiling) test facility under steam–water flow conditions. The test section is a vertical annulus of which the axial length is 4.165 m with a heater rod at the center of a channel. The inner and outer diameters of the test section and the heater rod are 35.5 mm and 9.98 mm, respectively. The test was performed by a two-stage approach. Stage-I for the measurement of local bubble parameters has been already done (Yun et al., 2009). The present work focused on the stage-II test for the measurement of local liquid parameters such as a local liquid velocity and a liquid temperature for a given flow condition of stage-I. A total of six test cases were chosen by following the test matrix of stage-I. The flow conditions are in the range of the heat flux of 370–563 kW/m², mass flux of 1110–2100 kg/(m² s) and inlet subcooling of 19–31 °C at pressure condition of 0.15–0.2 MPa. From the test, local liquid parameters were measured at 6 elevations along the test section and 11 radial locations of each elevation in addition to the previously obtained local void fraction, interfacial area concentration, Sauter mean diameter and bubble velocity. The present subcooled boiling (SUBO) data completes a data set for use as a benchmark, validation and model development of the Computational Fluid Dynamics (CFD) codes or existing safety analysis codes.     

Trapping of deuterium dissolved in fluidized Li by Y

Recovery of D dissolved in a liquid Li flow at low D concentration is experimentally investigated using a Y metal absorber under the two fluidized conditions: (a) in a vertical cylindrical tube and (b) in an agitated vessel. The target concentration is 1 appm in Li around at 300 °C. The two concentrations of D remaining in Li and recovered by Y are detected by a dissolution method using H₂O with depleted-D and HNO₃. The main released species is HD. A small amount of HDO released is reduced to HD by a Mg particle bed. It is found that HF-treated Y can absorb H isotopes at the target temperature and concentration. The chemical dissolution technique is found to be useful to specify the two absolute concentrations of D recovered by Y and D remaining in Li.     

Effect of transient heating loads on beryllium

Beryllium will be used as a plasma facing material for ITER first wall. It is expected that erosion of beryllium under transient plasma loads such as the edge-localized modes (ELMs) and disruptions will mainly determine a lifetime of ITER first wall. The results of recent experiments with the Russian beryllium of TGP-56FW ITER grade on QSPA-Be plasma gun facility are presented. The Be/CuCrZr mock-ups were exposed to upto 100 shots by deuterium plasma streams with pulse duration of 0.5 ms at ∼250 °C and average heat loads of 0.5 and 1 MJ/m². Experiments were performed at 250 °C. The evolution of surface microstructure and cracks morphology as well as beryllium mass loss are investigated under erosion process.     

The effect of displacement damage on deuterium retention in ITER-grade tungsten exposed to low-energy,high-flux pure and helium-seeded deuterium plasmas

Samples prepared from polycrystalline ITER-grade tungsten were damaged by irradiation with 20 MeV W ions at room temperature to a fluence of 1.4 × 10¹⁸ W/m². Due to the irradiation, displacement damage peaked near the end-of-range, 1.35 μm beneath the surface, at 0.89 displacements per atom. The damaged as well as undamaged W samples were then exposed to low-energy, high-flux (10²² D/m² s) pure D and helium-seeded D plasmas to an ion fluence of 3 × 10²⁶ D/m² at various temperatures. Trapping of deuterium was examined by the D(³He,p)⁴He nuclear reaction at ³He energies varied from 0.69 to 4.0 MeV allowing determination of the D concentration at depths up to 6 μm. It has been found that (i) addition of 10% helium ions into the D plasma at exposure temperatures of 440–650 K significantly reduces the D concentration at depths of 0.5–6 μm compared to that for the pure plasma exposure; (ii) generation of the W-ion-induced displacement damage significantly increases the D concentration at depths up to 2 μm (i.e., in the damage zone) under subsequent exposures to both pure D and D–He plasmas.     

Molecular dynamics simulations of argon cluster impacts on a nickel film surface

In this study we probe the surface phenomena that occur on nickel thin films after argon cluster impacts by performing several simulations using various energies. The simulations are carried out based on a molecular dynamics (MD) approach. The argon cluster consists of 353 atoms with energies ranging from 1 keV to 3.0 keV. The simulation results show that when the incident energy is 1 keV, the surface retains its smoothness after impact although a slight thermal effect appears near the surface beneath the impact area. Increasing the argon cluster energy to 2 keV causes the atoms in the film to shift slightly under impact and a small hillock appears on the film surface after impact. When the cluster energy increases to 3 keV, a hemispherical crater will appear on the film surface after impact. In addition, a shock wave is generated within the film due to the impact, which propagates toward to the substrate in a hemispherical shape. These shock wave related phenomena are difficult to probe experimentally on an atomic level however molecular dynamics simulations are a suitable tool for investigating the shock wave phenomena in thin film.     

Fatigue cracking of a bare steel first wall in an inertial confinement fusion chamber

Inertial confinement fusion power plants will deposit high energy X-rays onto the outer surfaces of the first wall many times a second for the lifetime of the plant. These X-rays create brief temperature spikes in the first few microns of the wall, which cause an associated highly compressive stress response on the surface of the material. The periodicity of this stress pulse is a concern due to the possibility of fatigue cracking of the wall. We have used finite element analyses to simulate the conditions present on the first wall in order to evaluate the driving force of crack propagation on fusion-facing surface cracks.Analysis results indicate that the X-ray induced plastic compressive stress creates a region of residual tension on the surface between pulses. This tension film will likely result in surface cracking upon repeated cycling. Additionally, the compressive pulse may induce plasticity ahead of the crack tip, leaving residual tension in its wake. However, the stress amplitude decreases dramatically for depths greater than 80–100 μm into the fusion-facing surface. Crack propagation models as well as stress-life estimates agree that even though small cracks may form on the surface of the wall, they are unlikely to propagate further than 100 μm without assistance from creep or grain erosion phenomena.     

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/m² 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.     

Operational status of the Magnum-PSI linear plasma device

The construction phase of the linear plasma generator Magnum-PSI at the FOM institute DIFFER has been completed and the facility has been officially opened in March 2012. The scientific program to gain more insight in the plasma–wall interactions relevant for ITER and future fusion reactors has started.In Magnum-PSI, targets of a wide range of materials and shapes can be exposed to high particle, high heat flux plasmas (>10²⁴ ions m^?2 s^?1; >10 MW/m²). For magnetization of the plasma, oil-cooled electromagnets are temporarily installed to enable pulsed operation until the device is upgraded with a superconducting magnet. The magnets generate a field of up to 1.9 T close to the plasma source for a duration of 6 s. Longer exposure times are available for lower field settings.Plasma characterizations were done with a variety of gases (H, D, He, Ne and Ar) to determine the machine performance and prepare for subsequent scientific experiments. Thomson scattering and optical emission spectroscopy were used to determine the plasma parameters while infrared thermography and target calorimetry were used to determine the power loads to the surface.This paper reports on the status of Magnum-PSI and its diagnostic systems. In addition, an overview of the plasma parameters that can be achieved in the present state will be given.     

Flow and heat transfer characteristics of helium cooled IFMIF High Flux Test Module

The International Fusion Materials Irradiation Facility (IFMIF) is designated to generate a materials irradiation database for the future fusion reactors. The High Flux Test Module (HFTM) is located behind the lithium target where a structural damage rate of more than 20 dpa/fpy in steel specimen will be achieved in a volume of about 0.5 l. This paper concerns the thermo-hydraulic simulations of the HFTM with Ansys-CFX. The simulation domain consists of the whole HFTM with short fluid inlet and outlet sections. The turbulence models were assessed by comparing the simulations with in-house annular channel experiments. The flow distributions and heat transfer characteristics among various HFTM sub-channels are the main focuses of this study.     

Thermal–hydraulic and thermo-structural analysis of first wall for Indian DEMO blanket module

The first wall (FW) is one of the most important components of any fusion blanket design. India has developed two concepts of breeding blanket for the DEMO reactor: the first one is Lead–Lithium cooled Ceramic Breeder (LLCB), and the second one is Helium-Cooled Ceramic Breeder (HCCB) concept. Both the concept has the same kind of FW structure. Reduced Activation Ferritic Martensitic steel (RAFMS) used as the structural material and helium (He) gas is used to actively cool the FW structure. Beryllium (Be) layer of 2 mm is coated on the plasma side of the FW as the plasma facing material. Cooling channels running in radial–toroidal–radial direction in the RAFMS structure are designed to withstand the maximum He pressure of 8 MPa. Heat transfer coefficients (HTC) obtained form the correlations revealed that required cooling could be achieved by artificially roughened surface towards the plasma-side wall of He cooling channel which helps to keep the RAFMS temperatures below the allowable limit. A 1D analytical and 2D thermal–hydraulic simulation studies using ANSYS has been performed based on the heat load obtained from neutronics calculations to confirm the heat removal and structural integrity under various conditions including ITER transient events. The required helium flow through the cooling channels are evaluated and used to optimize the suitable header design. The detail design of FW thermal–hydraulics, thermo-structural analyses, and He flow distribution network will be presented in this paper.