Experimental study of onset of subcooled annular flow boiling

An experimental study on the onset of nucleate boiling (ONB) is performed for water annular flow to provide a systematic database for low pressure and velocity conditions. A parametric study has been conducted to investigate the effect of pressure, inlet subcooling, heat and mass flux on flow boiling. The test section includes a Pyrex tube with 21 mm inner diameter and a stainless steel (SS-304) rod with outer diameter of 6 mm. Pressure, heat and mass flux are in the range of 1.73 < P < 3.82 bar, 40 < q < 450 kW/m² and 70 < G < 620 kg/m² s, respectively. The results illustrate that inception heat flux is extremely dependent on pressure, inlet subcooling temperature and mass flux; for example in pressure, velocity and inlet subcooling as 3.27 bar, 230 kg/m² s and 41.3 °C; consequently q_w,ONB is 177.3 kW/m². In other case with higher inlet temperature of 71.5 °C and with P, 3.13 bar and G, 232 kg/m² s the inception heat flux reached to 101.6 kW/m². The data of ONB heat flux are over estimated from the existing correlation, and maximum deviation of wall superheat (ΔT_w,ONB) from correlations is 30%. Experimental data of inception heat flux are within 22% of that predicted from the correlation.     

High heat flux testing of mock-ups for a full tungsten ITER divertor

In order to evaluate the option to start the ITER operation with a full tungsten (W) divertor, the EU-DA launched an extensive R&D program. It consisted in its initial phase in the high heat flux (HHF) testing of W mock-ups and medium scale prototypes up to 20 MW/m² in the AREVA FE 200 facility (F). Critical heat flux (CHF) experiments were carried out on the items which survived the above thermal fatigue testing.After 1000 cycles at 10 MW/m², the full W Plasma Facing Components (PFCs) mock-ups successfully sustained either 1000 cycles at 15 MW/m² or 500 cycles at 20 MW/m².However, some significant surface melting, as well as the complete melting of a few monoblocks, occurred during the HHF thermal fatigue testing program representative of the present ITER requirements for the strike point region, namely 1000 cycles at 10 MW/m² followed by 1000 cycles at 20 MW/m².The results of the CHF experiments were also rather encouraging, since the tested items sustained heat fluxes in the range of 30 MW/m² in steady-state conditions.     

Manufacturing and testing of a HETS module for DEMO divertor

The development of a divertor concept for fusion power plants that is able to grant efficient recovery and conversion of the considerable fraction (～15%) of the total fusion thermal power incident is deemed to be an urgent task to meet in the EU Fast Track scenario. The He-cooled conceptual divertor design is one of the possible candidates. Helium cooling offers several advantages including chemical and neutronic inertness and the ability to operate at higher temperatures and lower pressures than those required for water cooling. The HETS (high-efficiency thermal shield) concept, initially developed by ENEA for water, has been adapted for use with He as coolant. This DEMO divertor concept is based on elements joined in series and protected by a hemispheric dome; it allows an increase of thermal exchange coefficient both for high speed of gas and for “jet impingement” effects of gas coming out from the internal side of hemispheric dome. It has been calculated to be capable of sustaining an incident heat flux of 10 MW/m² when operating at 10 MPa, an inlet He temperature of 600 °C, and an outlet temperature of 800 °C. The presented activity, performed in the frame of EFDA-TW5TRP-001 task, was focused on the manufacturing of a single HETS module and on its thermal–hydraulic testing. The materials used for the HETS module manufacturing were all DEMO-compatible: W for the armor material and the hemispherical-dome, DENSIMET for the exchanger body. The testing is performed by connecting the module to HEFUS3 He loop system that is a facility able to supply the He flow to the required testing conditions: 400 °C, 4–8 MPa and 20–40 g/s. The needed incident heat flux is obtained by RF inducting equipment coupled to an inductor coil installed just over the HETS module. A CFD analysis by ANSYS-CFX was performed in order to predict the thermal–mechanical behavior of the module and a final comparison with the experimental data is required to validate the CFD results. All parameters are monitored and recorded by data acquisition system.     

Design of a water cooled monoblock divertor for DEMO using Eurofer as structural material

The performed investigation focus on a monoblock type design for a water cooled DEMO divertor using Eurofer as structural material. In 2013, a study case of such a concept was presented. It was shown that basic concepts using Eurofer as structural material are limited to an incident heat flux of 8 MW m⁻². Since, the EFDA agency issued new specifications. In this study, the conceptual design is reassessed with regard to specifications. Then, steady state thermal analyses and thermo-mechanical elastic analyses have been performed to define an upgrade of the geometry taking into account new specifications, design criteria and the maximum heat flux requirement of 10 MW m⁻². An analysis of the influence of each adjustable geometrical parameter on thermo-mechanical design criteria was performed. As a consequence, geometrical parameters were set in order to fit to materials requirements. For defined hydraulic conditions taken in the most favourable configuration, the limit of this design is estimated to an incident heat flux of 10 MW m⁻². Margin to critical heat flux and rules against progressive deformation/ratcheting in structural material limit the design.     

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 5 MW/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 = 1–3 m/s, P = 2–3 MPa, T = 110–170 °C). The heat flux was gradually varied in the range of 1–10 MW/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 ≥2 m/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 12 mm × 12 mm × 6 to 50 mm × 50 mm × 8 mm) obtained by the HHF (variable in the range of 3–8 MW/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.     

Model validation of GAMMA code with heat transfer experiment for KO TBM in ITER

By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) test blanket module (TBM) for testing in ITER. A performance analysis for the thermal–hydraulics and a safety analysis for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs multicomponent mixture analysis), which was developed by the gas cooled reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM first wall (FW) mock-up made from the same material as the KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 1.1, 1.9 and 2.9 MPa, and under various ranges of flow rate from 0.0105 to 0.0407 kg/s with a constant wall temperature condition. In the present study, a thermal–hydraulic test was performed with the newly constructed helium supplying system, in which the design pressure and temperature were 9 MPa and 500 °C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 3 MPa pressure, 30 °C inlet temperature and 70 m/s helium velocity, which are almost same conditions of the KO TBM FW. One side of the mock-up was heated with a constant heat flux of 0.3–0.5 MW/m² using a graphite heating system, KoHLT-2 (Korea heat load test facility-2). Because the comparison result between CFX 11 and GAMMA showed a difference tendency, the modification of heat transfer correlation included in GAMMA was performed. And the modified GAMMA showed the strong parity with CFX 11 calculation results.     

Experimental study on critical heat flux on a downward-facing stainless steel disk in confined space

Critical heat flux (CHF) is experimentally studied on a relatively large downward-facing surface with a heated stainless steel disk diameter of D = 300 mm in confined space at atmospheric pressure using water as the working fluid. The bulk working fluid is subcooled. The gap size s can be adjusted to 0.9, 2.2, 2.6, 3.0, 3.2, 5.0, 7.0, 10.0, 13.0, 15.6, 19.5, 25.0, 36.0, 51.0 and 77 mm. We found that the average CHF under the present condition is approximately 0.25 MW/m² which is only about 23% of which occurs on an upward-facing surface without confined space in pool boiling. The CHF increases with the increase of the gap size when the gap size is smaller than 7 mm and it is a function of Bond and Jakob numbers when the gap size is larger than 7 mm.     

Computational studies of thermoelectric MHD driven liquid lithium flow in metal trenches

The LiMIT system (Lithium/Metal Infused Trenches) is an innovative plasma-facing component for tokamak divertors, recently proposed at the University of Illinois. Thanks to the coupling of two metals having different Seebeck coefficients, the device is able to generate internal thermoelectric currents as a response to an incoming heat flux from the plasma. One of the two metals is liquid lithium and the second metal is a solid composing the trenches (tungsten, or molybdenum, or stainless steel, etc.). Together with the high toroidal magnetic field, the liquid lithium is propelled by a JxB electrodynamic force inside the solid trenches. In the present work we present a numerical characterization of the device. The diffusion–advection of heat is solved together with the Navier–Stokes equations forced by the JxB electrodynamic force, comprising the thermoelectric contribution. We report parametric plots to show the influence of the toroidal magnetic field and of the plasma heat flux. It is found that the average flow velocity of the liquid lithium peaks at a critical magnetic field, always lower than 1.0 T, and then decreases with an inverse law in the range of tokamak-relevant fields. The flow velocity of the lithium increases with a square-root law versus an increasing heat flux. The heat transfer coefficient of the cooling channels is parametrically investigated, revealing that coefficients higher than >4000 W/m² K are needed for the device in order to withstand heat fluxes of 10 MW/m².     

Numerical and experimental studies of Electra's scalloped transmission foil cooling with small impinging jets

The 5 Hz rep-rate operation of the Electra KrF laser necessitates the cooling and protection of the transmission foil that is subject to the pulsating electron beam bombardment. The pulsed volumetric heating from the e-beam attenuation heats up the foil (～2.54 × 10^?5 m thick) rapidly and often causes the foil to fail, increasing the operation cost and down time for the laser. Various methods have been investigated forheat transfer enhancement. While elevated heat transfer was achieved, the previous methods assume a flat foil shape. The actual foil shape is scalloped due to the pressure difference across the foil during the laser operation. Also a new “scalloped” foil design was proposed for thermal stress reduction. This paper investigates the applicability of small locally impinging jets to cooling the scalloped-shaped foil. The jets were formed through a line of small circular openings on two stainless-steel jet tubes aligned with the foil edges having the two columns of jets impinging on the foil obliquely in a staggered pattern for improved coverage. CFD simulations were used to optimize jet configurations. Experiments were performed that utilize a scalloped foil strip which matched the foil shape between two neighboring supporting ribs in the Electra hibachi. Jet diameters and jet velocities were varied at a surface heat flux greater than 20.0 kW/m². Substantial heat transfer enhancement with impinging jets was observed. Average Nusselt numbers were correlated with jet Reynolds number and the normalized jet-to-foil distance. The study indicates that the impinging jets can effectively enhance heat transfer for the scalloped foil and can be a promising method for actual foil coolingof KrF lasers, including Electra.     

Status of engineering design of liquid lithium target in IFMIF-EVEDA

In International Fusion Materials Irradiation Facility (IFMIF), intense neutron flux (4.5 × 10¹⁷ n/m² s) with a peak energy of 14 MeV are produced by means of two deuteron beams with a total current of 250 mA and maximum energy of 40 MeV that strike a liquid Li target circulating in a Li loop. Major design requirement is to provide a stable Li jet at a speed of 10–20 m/s with a surface wave amplitude on the Li flow less than 1 mm for handling of an averaged heat flux of 1 GW/m² under a continuous 10 MW deuterium beam deposition. The target system consists of a target assembly, a replaceable back-plate, a Li main loop and a Li purification loop. In July 2007, Engineering Validation and Engineering Design Activities (EVEDA) started under Broader Approach. In this paper, status of the engineering design of the IFMIF Li target system performed in 2007/2008 is described. The future EVEDA tasks to develop the target system are also summarized.     

Design and thermal performance of an improved mechanically attached module for divertor plate of LHD

A mechanically attached divertor module with improved performance has been developed and partially installed in the large helical device (LHD). The advantage of the new design is to eliminate any metal bolts on the armor tile surface which leads to high Z impurity emission. The new module consists of a couple of armor tiles made of iso-graphite and a thin graphite sheet, which are tightened by two stainless steel (SS) or TZM bolts horizontally sandwiching a SS cooling pipe. The heat flows directly from the tile surface to the cooling pipe. The previous design used a copper heat sink has been used. Steady heat flux tests up to 1.2 MW/m² have been carried out for the new module without any trouble using a electron gun ACT while the allowable heat flux of the previous module was limited below 0.3 MW/m² due to deformation of the copper heat sink. The outgassing from the new module during heat flux tests up to 0.5 MW/m² is decreased to about one-third. Thermal fatigue test up to 115 cycles under a steady heat flux of 1 MW/m² for the new module has been performed without any troubles. Comparisons between the previous and new modules for the structure, thermal performance, and outgassing are made.     

Development of a high energy pulsed plasma simulator for the study of liquid lithium trenches

To simulate detrimental events in a tokamak and provide a test-stand for a liquid-lithium infused trench (LiMIT) device [1], a pulsed plasma source utilizing a theta pinch in conjunction with a coaxial plasma accelerator has been developed. The plasma is characterized using a triple Langmuir probe, optical methods, and a calorimeter. Clear advantages have been observed with the application of a coaxial plasma accelerator as a pre-ionization source. The experimental results of the plasma gun in conjunction with the existing theta pinch show a significant improvement from the previous energy deposition by a factor of 14 or higher, resulting in a maximum energy and heat flux of 0.065 ± 0.002 MJ/m² and 0.43 ± 0.01 GW/m². A few ways to further increase the plasma heat flux for LiMIT experiments are discussed.     

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.     

2-D temperature distribution and heat flux of PFC in 2011 KSTAR campaign

KSTAR has reached a plasma current up to 630 kA, plasma duration up to 12 s, and has achieved high confinement mode (H-mode) in 2011 campaign. The heat flux of PFC tile was estimated from the temperature increase of PFC since 2010. The heat flux of PFC tiles increases significantly with higher plasma current and longer pulse duration. The time-averaged heat flux of shots in 2010 campaign (with 3 s pulse durations and I_p of 611 kA) is 0.01 MW/m² while that in 2011 campaign (with 12 s pulse duration and I_p of 630 kA) is about 0.02 MW/m². The heat flux at divertor is 1.4–2 times higher than that at inboard limiter or passive stabilizer. With the cryopump operation, the heat flux at the central divertor is higher than that without cryopump. The heat flux at divertor is proportional to, of course, the duration of H-mode. Furthermore, a software tool, which visualizes the 2D temperature distribution of PFC tile and estimates the heat flux in real time, is developed.     

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.     

Experimental studies on single-phase flow and heat transfer in a narrow rectangular channel

The state-of-the-art studies on single-phase flow and hear transfer in narrow rectangular channels shows some difference in terms of the agreement with the conventional theory. To further make clear this issue, the experimental studies on single-phase flow and heat transfer in a narrow rectangular channel with deioned water as test fluid was carried out. The narrow rectangular channel had the following dimensions: depth (e) = 2 mm, aspect ratio (e/b) = 0.05, length to diameter ratio (L/D_h) = 300, and mean wall relative roughness (?/D_h) = 8.3 × 10^?4. The experiments were performed with the channel oriented uprightly. The parameters that were varied during the experiments included the mass flow rate, inlet temperature and heat flux.Based on the measured temperatures, mass flow rates, pressure drops and heat fluxes, the isothermal and non-isothermal friction factors and the local and mean Nusselt numbers have been calculated. The correlations for the isothermal friction factors and the mean Nusselt numbers have been developed, and have a satisfactory agreement with the conventional theory. Based on the property ratio method, the correlations for non-isothermal friction factors have been proposed, but the new exponent (m) for modifying variable-property effect need to be developed.     

Numerical investigation of heat transfer enhancement in ribbed channel for the first wall of DFLL-TBM in ITER

As an important component of Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM), the first wall (FW) must withstand and remove the heat flux from the plasma (q″ = 0.3 MW/m²) and high nuclear power deposited in the structure at normal plasma operation scenario of ITER. In this paper the transverse ribs arranged along the plasma facing inner wall surface were used to enhance the heat transfer capability. After the validation compared with empirical correlations the Standard k–ω model was employed to do the numerical simulation using FLUENT code to investigate the heat transfer efficiency and flow performance of coolant in the ribbed channel preliminarily. The perforation on the bottom of rib was proposed near the lower heat transfer area (LHTA) to improve the heat transfer performance according to results of analyses.     

Design and fabrication of a high-temperature helium regenerator

Refractory metallic foams can increase heat transfer efficiency in gas-to-gas and liquid metal-to-gas heat exchangers by providing an extended surface area for better convection, i.e. conduction into the foam ligaments providing a “fin-effect,” and by disruption of the thermal boundary layer near the hot wall and ligaments by turbulence promotion. In this article, we describe the design of a high-temperature refractory regenerator (closed-loop recuperator) using computational fluid dynamics (CFD) modeling of actual foam geometries obtained through computerized micro-tomography. The article outlines the design procedure from geometry import through meshing and thermo-mechanical analysis and discusses the challenges of fabrication using pure molybdenum and TZM. The foam core regenerator is more easily fabricated, less expensive and performs better than refractory flat plate-type heat exchangers. The regenerator can operate with a maximum hot leg inlet temperature of 900 °C and transfer 180 kW to the cold leg using 100 g/s helium at 4 MPa. Future high heat flux experiments on helium-cooled plasma facing components will utilize the high temperature and high pressure capabilities of this unique regenerator. Similar components will be required to adapt fusion power reactors to high-efficiency Brayton power conversion systems and enable operation of advanced divertor and blanket systems.     

Design prospect of remountable high-temperature superconducting magnet

The remountable (mountable and demountable repeatedly) high-temperature superconducting (HTS) magnet has been proposed for huge and complex superconducting magnets in future fusion reactors to fabricate and repair easily the magnet and access inner structural components. This paper summarizes progress in R&D activities of mechanical joints of HTS conductors in terms of the electrical resistance and heat transfer performance at the joint region. The latest experimental results show the low joint resistance, 4 nΩ under 70 kA current condition using REBCO HTS conductor with mechanical lap joint system, and for the cooling system the maximum heat flux of 0.4 MW/m² is removed by using bronze sintered porous media with sub-cooled liquid nitrogen. These values indicate that there is large possibility to design the remountable HTS magnet for fusion reactors.     

Thermal effects and component cooling in Magnum-PSI

Magnum-PSI is a linear plasma generator, built at the FOM-Institute for Plasma Physics Rijnhuizen. Subject of study will be the interaction of plasma with a diversity of surface materials. The machine is designed to provide an environment with a steady state high-flux plasma (up to 10²⁴ H⁺ ions/m² s) in a 3 T magnetic field with an exposed surface of 80 cm² up to 10 MW/m². Magnum-PSI will provide new insights in the complex physics and chemistry that will occur in the divertor region of the future experimental fusion reactor ITER and reactors beyond ITER. The conditions at the surface of the sample can be varied over a wide range, such as plasma temperature, beam diameter, particle flux, inclination angle of the target, background pressure and magnetic field. An important subject of attention in the design of the machine was thermal effects originating in the excess heat and gas flow from the plasma source and radiation from the target.