In the framework of the roadmap of the DEMO reactor design pursued by the EUROfusion Programme, R&D activities have been promoted for the technological development of Plasma Facing Components (PFCs). Dedicated research activity has been undertaken at ENEA to support the development of technological solutions for the monoblock-pipe joining in order to reduce the use of materials having high activation and/or a degradation under neutron irradiation. For this purpose, a preliminary brazing alloy screening was carried out: a total of seven brazing alloys were identified and tested (i.e. Gemco, Nicuman23, TiCuNi, CuTiZrNi and three alloys with different percentages of Cu and Ge). For each brazing alloy, a wettability test on joint base materials (i.e., W and Cu) was performed. Then, three samples were fabricated joining tungsten monoblocks, without Cu interlayer, on a W fiber-reinforced Cu composite cooling pipe; other three samples were realized joining W monoblocks, with and without Cu interlayer on standard ITER-grade CuCrZr pipes. Non-destructive Ultrasonic Testing (UT) examinations were performed on each sample and showed that the monoblocks surface was not fully attached to Wf-Cu pipes; as regard the samples with CuCrZr pipes, excellent results have been achieved both in the case with and without Cu interlayer. From the results, Gemco seems to be the most promising commercial alloy among the tested ones, thanks to its low amount of Nickel and the good joining capabilities.

In the framework of the activity of target development of the European DEMO divertor, ENEA carried out an extensive ultrasonic testing campaign on more than 66 tungsten monoblock mock-ups. The EU-DEMO target development activity concerns primarily the comparison between the reference solution for the divertor targets (ITER-like) and two other concepts (Thermal Break Interlayer, Functionally Graded Interlayer), in which the interlayer between monoblocks and tube has been modified with the aim of increasing the component performance. In the same activity, many other aspects were also considered such as analyzing the influence of the tungsten monoblock supplier, interlayer thickness and interlayer manufacturing process (casting or diffusion bonding). Mock-ups were provided by ENEA (ITER-Like), CEA (Thin Graded Interlayer) and CCFE (Thermal Break Interlayer) and tested to thermal fatigue by high heat flux in the GLADIS facility in IPP - Garching center under DEMO relevant conditions. In this work the comparison between UT results obtained before and after the high heat flux test for the more significant mock-ups for phase 2 are carried out. After testing, mock-ups were cut in order to conduct post-mortem analyzes. Substantial agreement is found with UT results.

The divertor is one of the most challenging components for DEMO reactor both from the design and fabrication technology point of view, since it must be capable to withstand the high heat fluxes (HHF) expected during normal operation (up to 10 MW/m(2)) and slow transient events (up to 20 MW/m(2)), like loss of plasma detachment. Within the frame of the EUROfusion Consortium the "Target development" subproject inside the Work Package Divertor' (WPDIV) has been dedicated to achieve this performance studying different concepts. ENEA focused on the "ITER-like" target that consists of applying the ITER design and fabrication technology to DEMO targets. The ITER-like concept mock-up was at first designed and optimized by FE analysis then manufactured and checked by non-destructive testing (NDT), finally thermal fatigue testing (TFT) and destructive testing (DT) were performed too. This paper reports the comparison of the thermal behavior of the ITER-like mock-ups between the analysis performed using ANSYS-CFX (Computational Fluid Dynamic analysis, CFD) and the outcome of the High Heat Flux tests (HHF) made at GLADIS facility. Finally the mechanical behavior of the mock-ups has been compared in terms of ratchetting applying dedicated criteria of the ITER SDC-IC and the fatigue lifetime has been estimated by means of Low Cycle Fatigue Curves (LCFC).

One of the most critical parts of a high heat flux plasma facing component of a Tokamak divertor is the armour to heat sink joint. R&D activity has been launched in the frame of the EUROFUSION Power Plant Physics & Technology programme and in particular in the Divertor project area in order to investigate the possibility of using the International Thermonuclear Experimental Reactor (ITER) design and fabrication technology in an EU-DEMO tokamak. ENEA is being involved in the R&D ITER activities since several years developing suitable joining technologies for divertor plasma facing components fabrication. In fact, Hot Radial Pressing technique was used for the manufacturing of several small and medium scale prototypes that have been successfully tested by e-beam thermal fatigue testing. A new furnace was designed and installed at ENEA labs with the objective to extend the Hot Radial Pressing (HRP) manufacturing technology to EU-DEMO relevant divertor plasma facing components. This new furnace uses the pipe heat sink of these components as heater, speeding up the heating phase and allowing a better control of the pipe temperature. Considering that in EU-DEMO the thermo-hydraulic operative conditions of these components are different from ITER a FEA parametric optimization was performed and the geometry of the small mock-up to be manufactured was defined accordingly. The six manufactured mock-ups were examined by Ultrasonic and SATIR non-destructive examination methods in order to check the quality of the joining before their delivering to the high heat flux tests facility GLADIS for the thermal fatigue testing at high heat flux.

The vertical targets of the ITER divertor consist of high flux units (HFU) actively cooled: CuCrZr tubes armoured by tungsten and carbon/carbon fibre composite (CFC). The armour is obtained with holed parallelepiped blocks, called monoblocks, previously prepared and welded onto the tubes by means diffusion bonding. The monoblock preparation consists in the casting of a layer of copper oxygen free (Cu OFHC) inside the monoblock hole.Each HFU is covered with more than 100 monoblocks that have to be joined simultaneously to the tube. Therefore, it is very important to individuate any defects present in the casting of Cu OFHC or at the interface with the CFC before the monoblocks are installed on the units. This paper discusses the application of non-destructive testing by ultrasound (US) method for the control of the joining interfaces between CFC monoblocks and Cu OFHC, before the brazing on the CrCrZr tube.In ENEA laboratory an ultrasonic technique (UT) suitable for the control of these joints with size and geometry according to the ITER specifications has been developed and widely tested. Real defects in this type of joints are, however, still hardly detected by UT. The CFC surface has to be machined to improve the mechanical strength of the joint. This results in a surface not perpendicular to the ultrasonic wave. Moreover, CFC is characterized by high acoustic attenuation of the ultrasonic wave and then it is not easy to get information regarding the Cu/CFC bonding. Nevertheless, the UT sharpness and simplicity pushes to perform some further study. With this purpose, a sample with artificial defects induced on the CFC/Cu interface during the Cu casting, has been manufactured and investigated, both by pulse-echo technique and by different techniques. The results obtained by the UT on this sample are reported and compared with X-ray method. (C) 2013 Published by Elsevier B.V.

This paper discusses the application of non-destructive testing (NDT) by ultrasonic technique for the control of the joining interfaces of the ITER divertor vertical target plasma facing units. The defect detection capability has to be proved for both metal to metal and metal to carbon/carbon fibre composite (CFC) joints because these two types of joints have to be realized for the manufacturing of the high heat flux units. In this paper the UT results coming from the investigation performed during the manufacturing, but also after the thermal fatigue testing (up to 20MW/m2) of six mock-ups manufactured using the Hot Radial Pressure technology (HRP) in ENEA labs are presented and compared with the evidences from the final destructive examination. Regarding the Cu/CFC joint, the effectiveness of the ultrasonic test has been deeply studied due to the high acoustic attenuation of CFC to ultrasonic waves. To investigate the possibility to use the ultrasonic technique for this type of joint, an ‘ad hoc’ flat Cu/CFC joint sample, that reproduces the actual annular joint interfaces, was manufactured. This flat sample has the advantage of being easily tested by probes with different geometry and frequency. UT results are compared with X-ray and eddy current testing of the same sample.

The most critical part of a high heat flux (HHF) plasma facing component (PFC) is the armour to heat sink joint.An experimental study was launched by EFDA in order to define the acceptance criteria to be used for the procurements of the ITER Divertor PFCs.ENEA is involved in the European International Thermonuclear Experimental Reactor (ITER) R&D activities and together with Ansaldo Ricerche S.p.A. has manufactured several PFCs mock-ups using the Hot Radial Pressing and Pre-Brazed Casting technologies.According to the technical specifications issued by EFDA, ENEA and Ansaldo have collaborated to manufacture half of the samples with calibrated artificial defects required for this experimental study.After manufacturing, the samples were examined by ultrasonic and SATIR non-destructive examination (NDE) methods in order to confirm the size and position of the artificial defects.In particular, it was concluded that defects are detectable with these NDE techniques and they finally gave indication about the threshold of propagation during high heat flux experiments relevant with heat fluxes expected in ITER Divertor.This paper reports the manufacturing procedure used to obtain the required calibrated artificial defects in the CFC and W armoured samples as well as the NDE results and the thermal high heat flux results. (C) 2010 Elsevier B.V. All rights reserved.