Pulsed High Heat Flux Research Articles

Development and manufacturing of beryllium-armoring ITER enhanced heat flux FW toward series production in China

ITER’s enhanced heat flux (EHF) first wall (FW) provides thermal shielding to the other components behind it and limits its plasma boundary, consequently bearing high surface heat loads up to 4.7 MW m−2. China developed the EHF FW technologies in steady stages by making small mock-ups, semi-prototypes and a full-scale prototype (FSP), working toward series production at the Southwestern Institute of Physics. All the key technologies for bimetallic diffusion bonding, welding and assembly have been fully qualified. The Be armor tile size effects on thermal-fatigue performance have been evaluated using a pulsed high heat flux test with an EMS-400 electron beam facility. It was found that both the Be/CuCrZr bonding and its thermal-fatigue life are highly dependent on the tile size, and 12 × 12 mm2 Be tiles are acceptable for the required performance, either as individual tiles or in castellation by cutting larger ones. Small-scale mock-ups with such Be tiles survived 16 000 thermal cycles at 4.7 MW m−2, while EHF FW fingers with 16 × 16 mm2 Be tiles demonstrated unstable performance. In contrast, only BE tiles larger than 24 × 12 mm2 showed reliable diffusion bonding with the CuCrZr heat sink by hot isostatic pressing and consequently should be castellated into smaller ones. The manufacturing of the FSP finger goes through complex thermal cycles, using CuCrZr alloys in age-strengthening plus cold-rolling states, enabling it to maintain its tensile strength to the level of 285 MPa and its mean grain size less than 100 µm. The pipe welding for finger pairing and its assembly with the central beam are demonstrated, including enabling welding to be carried out twice for repair in a narrow space. The FSP showed good finger alignment, dimensional control and reliable vacuum tightness without any blockage of cooling channels.

Observation of surface deformation of tungsten exposed to single pulsed high heat flux and magnetic field for divertor design

Tungsten (W) is one of candidate surface materials for a plasma facing component (PFC) in a magnetically confinement fusion reactor such as a divertor from the perspective of its high melting point, high thermal conductivity, low tritium retention and low sputtering yield. In the fusion reactor in operating, the divertor surface is exposed by the transient heat load generated by plasma disruption or ELM. The heat load is sufficiently high to melt the W surface. Because of the repetition of melting and solidification, it is concerned that W surface deformation occurs, the droplets scatter and its heat resistance degrades. These are critical issues for developing the fusion reactor. In previous studies, the in-situ observations of the W specimen surface during irradiating by a pulsed laser simulating the heat load were conducted using a HSV camera. However, PFCs lies under intense magnetic field in the actual fusion-reactor condition. Magnetic field affects the motion of fluid by the effect of generated Lorentz force. Therefore, this study aims to clarify the melting and solidification behavior of W under the intense magnetic field. In order to simulate the heat load, the Nd/YAG laser, which has almost flatten spatial profile and rectangular temporal profile, was irradiated on a W specimen. The power density of the laser was 3.12 GW/m2 and 3.9 GW/m2 and the pulse width was 3 [ms]. Helmholtz-like electromagnetic coils were applied for generating pulsed magnetic field. The magnetic density was also changed in the range from 0 to 4 [T] in this study. From observation of the light emitted from the heated W surface using a high-speed video camera and measurement of irradiation traces, the behavior of molten W and the effect of the magnetic field will be reported and discussed.

In-situ measurement of surface modifications of tungsten exposed to pulsed high heat flux for divertor design in tokamak-type fusion nuclear reactors

Abstract A tungsten (W) divertor is exposed to pulsed high heat caused by disruptions and Type-I Edge Localized Modes (ELMs). Therefore, it is important to investigate surface morphology of W during phase change. Heretofore, W melting dynamics under disruption-like heat load has been researched and tried to modify the optical system of our apparatus for knowing more detailed surface behavior. In this paper, the two measurement methods are shown; the in-situ observation of the W surface modification by a High Speed Video (HSV) camera and 3-Dimensional (3D) measurement of melting W surface by stereoscopic photography technique (SPT). In the former, the detailed process of W melting was clarified by observing W surface modification and analyzing luminance at the center of laser-irradiated spot. It was confirmed that luminescence at the center area of heat load fluctuated violently due to the phase change of W. In the latter, 3D images of W surface shape during melting was obtained using SPT and the results indicated that the W surface fluctuated in melting and gathered to the center gradually in solidifying. The calibration accuracy was still scope for improvement but the 3D measurement of W surface behavior loaded by high heat flux was established successfully.

Brittle-ductile transition temperature of recrystallized tungsten following exposure to fusion relevant cyclic high heat load

The lifetime of tungsten (W) monoblocks under fusion conditions is ambivalent. In this work, the microstructure dependent mechanical behaviour of pulsed high heat flux (HHF) exposed W monoblock is investigated. Two different microstructural states, i.e. initial (deformed) and recrystallized, both machined from HHF exposed monoblocks are tested using tensile and small punch tests. The initial microstructural state reveals a higher fraction of low angle boundaries along with a preferred orientation of crystals. Following HHF exposure, the recrystallized state exhibits weakening of initial texture along with a higher fraction of high angle boundaries. Irrespective of the testing methodology, both the microstructural states display brittle failure for temperatures lower than 400∘C. For higher temperatures (>400∘C), the recrystallized microstructure exhibits more ductile behaviour as compared to the initial state. The observed microstructural state-dependent mechanical behaviour is further discussed in terms of different microstructural features. The estimated brittle-to-ductile transition temperature (BDTT) range is noticed to be lower for the recrystallized state as compared to the initial state. The lower BDTT in the recrystallized state is attributed to the high purity of the W in combination with its low defect density, thereby preventing segregation of impurities at the recrystallized boundaries and the related premature failure. Based on this observation, it is concluded that the common opinion of the aggravation of BDTT in W due to recrystallization is not unerring, and as a matter of fact, recrystallization in W could be instrumental for preventing the self-castellation of the monoblocks.

Shaping the Aperture of the Absolute Collimator in DIII-D Neutral Beamline

The absolute collimator currently in service at the DIII-D NB injection system has experienced localized melting and damage. As part of the DIII-D 210-deg beamline co-counter conversion, a new absolute collimator was needed, and the opportunity to resolve melting was found on the off-axis beamline configuration. The pulsed high heat flux and uneven distribution of the heat loads required the aperture surface to be axially extended to spread out and reduce the surface heat flux. Geometric sculpting of the absolute collimator aperture based on the baseline dimension was performed using ANSYS CFX software. The reshaped absolute collimator aperture surface reduces the impinged heat flux to below ~4 MW/m2. Two interchangeable inserts are designed to occupy the high heat flux region for mitigating the thermal-induced stresses. The design achieves the objective of 6-s pulse lengths with 10-min repetition rates using the original peripheral conduit cooling system in the new collimator with minor changes.

Study on melting and solidification behaviors of tungsten loaded by high heat flux for divertor in tokamak fusion reactor

Tungsten is a candidate as a surface material of divertor in Tokamak fusion reactor. A pulsed high heat flux emitted in plasma disruption and Type I Edge Localized Mode (ELM) is considered to melt the tungsten surface repeatedly. The repetitive phase change can transmute surface characteristics of divertor and then reduce the thermal resistance. For the safety design and the development acceleration of fusion demonstration reactor, investigation and understanding of surface melting and solidification procedure are important. In previous studies, experiments have been conducted by Ueda et al. and the surface after the laser irradiation had the geometry as a crater with a center cone. However, the formation mechanism of the crater was not fully understood. In this paper, two measurement methods are studied. One is an in-situ observation method of tungsten and molybdenum surface during the laser irradiation using high speed video camera, and another is a 3-dimensional measurement of tungsten surface shape after the laser irradiation experiments using a stereoscopic photography technique. In the former, it was confirmed that a molten pool had wavy surface pattern until solidification, and the applicability of the stereoscopic photography technique for the in-situ observation during the laser irradiation was indicated in the latter.

Surface morphology changes of tungsten exposed to high heat loading with mixed hydrogen/helium beams

We discuss the surface morphology modification of W samples observed after simultaneous heat and particle loading using a mixed H/He particle beam with a He concentration of 1at.%. The applied heat flux of 10MW/m2 is representative for the normal operation of the divertor of DEMO or a power plant. The long pulse high heat flux experiments on actively water-cooled W samples were performed in the GLADIS facility at surface temperatures between 600°C and 2000°C. This allows together with the applied total fluences between 1×1024m−2 and 1×1026m−2 the temperature- and fluence dependent study of the growing nano-structures. We analyse in detail the surface modifications up to a depth of several μm by scanning electron microscopy combined with focussed ion beam preparation. The hydrogen and helium release of the samples is analysed by long term thermal desorption spectroscopy and compared with the prediction of a diffusion trapping model.

Phase-field modeling of thermomechanical damage in tungsten under severe plasma transients

Tungsten is now a primary candidate for plasma facing components in fusion energy systems because of its numerous superior thermophysical properties. International efforts are currently focused on the development of tungsten surfaces that can intercept ionized plasma and pulsed high heat flux in magnetic fusion confinement devices. Thermal shock under transient operating conditions, such as edge localized modes, have experimentally been shown to lead to severe surface and sub-surface damage. We present here a computational multiphysics model to determine the relationship between the thermomechanical loading conditions and the onset of damage and failure of tungsten surfaces. The model is based on thermo-elasto-plasticity constitutive relations, and is developed within the framework of the phase-field method. A coupled set of partial differential equations is solved for the temperature, displacement, and a damage phase fields under severe plasma transient loads. The results clearly show the initiation and propagation of surface and sub-surface cracks as a result of the transient high heat flux. The severity of surface cracking is found to correlate primarily with the magnitude of the near-surface temperature gradient.

Recrystallization and grain growth behavior of rolled tungsten under VDE-like short pulse high heat flux loads

Short pulse heat loads expected for vertical displacement events (VDEs) in ITER were applied in the high heat flux (HHF) test facility GLADIS at IPP-Garching onto samples of rolled W. Pulsed neutral beams with the central heat flux of 23MW/m2 were applied for 0.5, 1.0 and 1.5s, respectively. Rapid recrystallization of the adiabatically loaded 3mm thick samples was observed when the pulse duration was up to 1.0s. Grains grew markedly following recrystallization with increasing pulse length. The recrystallization temperature and temperature dependence of the recrystallized grain size were also investigated. The results showed that the recrystallization temperature of the W grade was around 2480°C under the applied heat loading condition, which was nearly 1150°C higher than the conventional recrystallization temperature, and the grains were much finer. A linear relationship between the logarithm of average grain size (ln d) and the inverse of maximum surface temperature (1/Tmax) was found and accordingly the activation energy for grain growth in temperature evolution up to Tmax in 1.5s of the short pulse HHF load was deduced to be 4.1eV. This provided an effective clue to predict the structure evolution under short pulse HHF loads.

Thermo-mechanical analysis of the Wendelstein 7-X divertor

The stellarator Wendelstein 7-X (W7-X) has a divertor consisting of 10 units installed inside the plasma vessel (PV). It was decided not to install the long pulse high-heat flux (HHF) divertor targets at the first two years stage of W7-X operation and to start with an adiabatically cooled test divertor unit (TDU) and shorter plasma pulses operation. This allows to accumulate operation experience with much simpler components, and as a result to adjust accurately the actively cooled HHF divertor which replaces the TDU for the stationary operation. Finite element (FE) analyses have been performed for better understanding of thermo-mechanical problems of divertor targets, and to guide the design of the TDU and HHF divertors. This paper presents the detailed results of the temperature response, the deformation and thermal stress of the divertor components.

Progress in the design and development of a test divertor (TDU) for the start of W7-X operation

With the aim of starting plasma operation in 2014 the Wendelstein 7-X (W7-X) project has developed a scenario with an intermediate operational phase in which a reduced number of in-vessel components are installed prior to going to the fully equipped, full performance phase. An important part of this scenario is the test divertor (TDU), an adiabatically cooled, reduced pulse length device that will allow machine operation over a comparable range of plasma configurations as the final long pulse high-heat flux (HHF) divertor, which will be installed after two years operation of the TDU. The design and development of the TDU has started by clearly defining the requirements of the TDU with regard to performance, material selection, geometry, interfaces to other components, diagnostic integration, installation requirements and tolerances and quality management. In fact, the whole TDU development will follow the procedures laid down for design and development of components within the W7-X project. This paper provides a description of the design and development of the TDU to date.

Thermal performance and 3D analysis of advanced mechanically joined divertor plate for LHD under steady state high heat flux

Thermal performances of an advanced mechanically joined module (MJM) under steady state high heat fluxes of 2.0–4.25 MW/m 2 and thermo-mechanical analysis are described. The advanced MJM was designed to apply for a helical divertor plate of a large helical device (LHD) at the next experimental phase II. The advanced MJM has a unified armor/heat sink made of carbon/carbon composite different from the normal MJM with a separated copper heat sink. To evaluate the thermal performance of the advanced MJM, short pulse high heat flux test up to 5.4 MW/m 2 and steady state high heat flux tests up to 4.25 MW/m 2 have been carried out using a test facility ACT. Moreover, the thermal fatigue test of the advanced MJM up to 150 cycles under steady state high heat flux of 2.5 MW/m 2 has been performed. After theses tests, no apparent damage and no cracking on the armor tile were observed although there was a little increase in the armor/heat sink temperature during the thermal fatigue test. To evaluate the thermal stress in the armor/heat sink of the MJM during steady state high heat flux test and to optimize the structure, thermo-mechanical analyses are done using a 3D CAD.

Loading nature of the interfacial cracks in a joint component under fusion-relevant thermal loads

One of the standard design concepts for divertor components in a fusion reactor is the bonded joint structure. Understanding the loading nature of interfacial cracks are significant for the assessment of structural integrity of divertor joint components. In this paper, the thermomechanical loading nature of interfacial cracks is discussed. A bi-material joint element consisting of the CFC/TZM system is considered. A typical fusion operation condition is simulated assuming a pulsed high heat flux loading. Stress singularities near the interfacial crack tips are characterized quantitatively in terms of the fracture mechanical parameters. The evolution of the stress intensity factors and the energy release rate during the given transient thermal load are determined. The difference in loading characteristics between the edge crack and the center crack is discussed. High heat flux cycling tests are performed on brazed CFC/TZM divertor elements in an electron beam test facility. The microstructures of the damaged interface agree with the predicted fracture modes. The loading nature and possible failure mechanisms are discussed for a fusion-relevant thermal loading.

Initiation of Melting and Evaporating of Materials under Plasma Disruption.

Abstract In a tokamak reactor, a pulsed high heat flux, whose heat flux should result in up to 2x105MW/m2, would be injected to the plasma facing components (PFCs) due to a plasma disruption. By this heat load the PFCs are melted, evaporated and cracked. In this study melting and evaporation behavior was investigated through performing the experiment with use of an electron beam to simulate heat loads due to plasma disruptions and the numerical and approximate analyses. Experimental and analytical results show that the evaporated mass increases abruptly at a curtain heat flux and time, and that the thickness of melting layer has the maximum at a curtain heat flux. From dimensionless expressions of such results the simple dimensionless correlations between the relevant values were obtained to estimate the significant characteristics such as evaporation thresholds and maximum melting layer thickness. Applying them to evaluation of the thermal resistance of materials against various heat loads due to a plasm...

Pulse high heat flux experiment with laser beams on graphite

Pulse high heat flux experiment was performed on a first wall candidate graphite material using a pulse laser beam generator. Real-time detection of gases and particles emitted from the specimen was carried out with a quadrupole mass analyser (QMA). Various kinds of hydrocarbon compounds were detected during and after the laser shots. Emission of carbon clusters, indicating sublimation from the surfaces, was also observed. The behavior of these gases and particles in vacuum, the effects of laser irradiation condition on gas/particle emission and surface morphological changes were investigated.