Non-destructive examination of the bonding interface in DEMO divertor fingers

Plasma facing components (PFCs) with tungsten (W) armor materials for DEMO divertor require a high heat flux removal capability (at least 10 MW/m² in steady-state conditions). The reference divertor PFC concept is a finger with a tungsten tile as a protection and sacrificial layer brazed to a thimble made of tungsten alloy W – 1% La₂O₃ (WL10). Defects may be located at the W thimble to W tile interface. As the number of fingers is considerable (>250,000), it is then a major issue to develop a reliable control procedure in order to control with a non-destructive examination the fabrication processes. The feasibility for detecting defect with infrared thermography SATIR test bed is presented. SATIR is based on the heat transient method and is used as an inspection tool in order to assess component heat transfer capability. SATIR tests were performed on fingers integrating or not the complex He cooling system (steel cartridge with jet holes). Millimeter size artificial defects were manufactured and their detectability was evaluated. Results of this study demonstrate that the SATIR method can be considered as a relevant non-destructive technique examination for the defect detection of DEMO divertor fingers.     

Drag force on a circular cylinder midway between two parallel plates at very low Reynolds numbers—Part 1: Poiseuille flow (numerical)

At very low Reynolds numbers, we calculate the drag force exerted on a circular cylinder in cross flow fixed midway between two parallel plane walls which are fixed while the fluid experiences a Poiseuille profile at upstream and downstream. The drag wall correction factor is numerically investigated from a very weak interaction to the lubrication regime. The Navier-Stokes and continuity equations are expressed in the stream function and vorticity formulation and are rewritten in an orthogonal system of curvilinear co-ordinates. These equations are solved with using a finite differences method. The generation of the grid was carried out by the singularities method. We calculated the separate contributions of the pressure and viscous forces numerically. At very weak interactions, our numerical results are in good agreement with those obtained analytically by Harrison (Trans. Camb. Phil. Soc. 23 (1924) 71) and Faxèn (Proc. Roy. Swed. Acad. Eng. Sci. 187 (1946) 1). In the lubrication regime these numerical calculations are in very good agreement with those we carried out by asymptotic expansion. So that, the accuracy of the numerical code is tested. This analysis allowed us to show how that the pressure term prevails over the viscosity term in the lubrication regime. At very weak interaction, these forces have the same value.     

Potential and limits of water cooled divertor concepts based on monoblock design as possible candidates for a DEMO reactor

In this paper water-cooled divertor concepts based on tungsten monoblock design identified in previous studies as candidate for fusion power plant have been reviewed to assess their potential and limits as possible candidates for a DEMO concept deliverable in a short to medium term (“conservative baseline design”). The rationale and technology development assumptions that have led to their selection are revisited taking into account present factual information on reactor parameters, materials properties and manufacturing technologies.For that purpose, main parameters impacting the divertor design are identified and their relevance discussed. The state of the art knowledge on materials and relevant manufacturing techniques is reviewed. Particular attention is paid to material properties change after irradiation; phenomenon thresholds (if any) and possible operating ranges are identified (in terms of temperature and damage dose). The suitability of various proposed heat sink/structural and sacrificial layer materials, as proposed in the past, are re-assessed (e.g. with regard to the possibility of reducing peak heat flux and/or neutron radiation damages). As a result, potential and limits of various proposed concepts are highlighted, ranges in which they could operate (if any) defined and possible improvements are proposed.Identified missing point in materials database and/or manufacturing techniques knowledge that should be uppermost investigated in future R&D activities are reported.This work has been carried out in the frame of EFDA PPPT Work Programme activities.     

Design optimization of plasma facing component with functional gradient material Cu/W interlayer

In the current water cooled divertor concept, tungsten is an armor material and CuCrZr is a structural material. The difference in thermal expansion coefficient between tungsten and copper alloy causes a thermal mismatch between them resulting in accumulation of stresses which would yield failure of the joint. To reduce such stresses, a functionally graded material or a graded layer is introduced between tungsten and copper alloy. A fabrication via a powder metallurgy process such as SPS (Spark Plasma Sintering) is proposed. In this work, an actively cooled component composed of a functionally graded W/Cu is simulated under high heat flux using the finite element modeling method. Thermo-mechanical behaviors, due to the existence of temperature gradient, are analyzed by a modification of the layer compositions and the number of layers. Such behaviors are evaluated to determine the optimal water cooled divertor design. It appears that the copper concentration in the interlayer defines the mechanical response. We suggest, as a conclusion, to use three layers having an optimized copper concentration to fabricate them using SPS technology.     

The WEST project: PFC shaping solutions investigated for the ITER-like W divertor

Among major issues for PFCs design, the impact of leading edges (exposed surface) which would be directly intersected by particles following magnetic field lines at glancing incident angles in the high heat flux areas is much discussed. This paper presents the key outcome of a thermal analysis performed on different shaping solutions for the ITER-like W monoblocks occurred for the components of the WEST (W Environment for Steady state Tokamak) divertor which could shadow any direct leading edge and to counteract a potential misalignment due to assembly tolerance. The results, in terms of surface temperature rise and wall heat flux into the cooling channel, are discussed for magnetic field lines incident at glancing angles expected in the higher heat flux regions of divertor (i.e. close to the strike point regions) and for perpendicular incident heat flux up to 20 MW/m².     

Critical heat flux acoustic detection: Methods and application to ITER divertor vertical target monitoring

Actively cooled plasma facing components (PFCs) have to exhaust high heat fluxes from plasma radiation and plasma–wall interaction. Critical heat flux (CHF) event may occur in the cooling channel due to unexpected heat loading or operational conditions, and has to be detected as soon as possible. Therefore it is essential to develop means of monitoring based on precursory signals providing an early detection of this destructive phenomenon, in order to be able to stop operation before irremediable damages appear.Capabilities of CHF early detection based on acoustic techniques on PFC mock-ups cooled by pressurised water were already demonstrated. This paper addresses the problem of the detection in case of flow rate reduction and of flow dilution resulting from multiple plasma facing units (PFU) which are hydraulically connected in parallel, which is the case of ITER divertor. An experimental study is launched on a dedicated mock-up submitted to heat loads up to the CHF. It shows that the measurement of the acoustic waves, generated by the cooling phenomena, allows the CHF detection in conditions similar to that of the ITER divertor, with a reasonable number of sensors. The paper describes the mock-ups and the tests sequences, and comments the results.     

Improvement of non destructive infrared test bed SATIR for examination of actively cooled tungsten armour Plasma Facing Components

For steady state (magnetic) thermonuclear fusion devices which need large power exhaust capability and have to withstand heat fluxes in the range 10–20 MW m^?2, advanced Plasma Facing Components (PFCs) have been developed. The importance of PFCs for operating tokamaks requests to verify their manufacturing quality before mounting. SATIR is an IR test bed validated and recognized as a reliable and suitable tool to detect cooling defaults on PFCs with CFC armour material. Current tokamak developments implement metallic armour materials for first wall and divertor; their low emissivity causes several difficulties for infrared thermography control. We present SATIR infrared thermography test bed improvements for W monoblocks components without defect and with calibrated defects. These results are compared to ultrasonic inspection. This study demonstrates that SATIR method is fully usable for PFCs with low emissivity armour material.     

Fatigue lifetime of repaired high heat flux components for ITER divertor

Thermal fatigue behaviour of repaired monoblocks was assessed from High Heat Flux (HHF) tests up to 20 MW m^?2 on 11 components. Among these components, 8 monoblocks were repaired (2 CFC and 6 tungsten). These components were manufactured by two EU industries: ANSALDO Ricerche and PLANSEE. Non destructive examination was performed on SATIR thermography test bed before and after HHF tests. SATIR results show that repaired monoblocks have a good thermal exhaust capability before HHF tests. For all monoblocks, no degradation of thermal properties was noticed during cycles at 10 MW m^?2. After hundreds of cycles at 20 MW m^?2, two W repaired monoblock melted. Post-HHF SATIR examination revealed a degradation of thermal properties which is systematic for W melted monoblocks and non-systematic for W repaired ones. For CFC repaired monoblocks, no damage was observed up to 20 MW m^?2. For the first ITER divertor set, specifications for the pre-qualification are that CFC (Resp. W) components have to sustain in steady state 1000 cycles at 10 MW m^?2 (Resp. 3 MW m^?2) followed by 1000 cycles at 20 MW m^?2 (Resp. 5 MW m^?2). For the first ITER divertor set, the repair process is validated for CFC and W monoblocks.