G. Dell'Orco

European achievements for ITER high heat flux components

This paper summarises the main activities carried out by the EU Home Team to develop suitable solutions for the ITER high heat flux components, namely the divertor, the baffle and the limiter. The available results demonstrate that the EU have the capability to manufacture high heat flux components with carbon fibre reinforced carbon, tungsten and beryllium armours which all exceed the ITER design requirements.

European development of prototypes for ITER high heat flux components

The extensive EU research and development, on international thermonuclear experimental reactor (ITER) high heat flux (HHF) components aims at the demonstration of prototypes for the divertor and baffle with challenging operating requirements. The recent progress of this development is summarised in the paper, particularly concerning the manufacture and testing of mock-ups and prototypes. The available results demonstrate the feasibility of robust solutions with carbon and tungsten armour.

Thermal–mechanical test on ITER primary first wall mock-ups

Abstract In 1998, in the frame of the ITER EDA phase, an European R&D Programme for the Blanket Design was implemented for developing and selecting the materials and the relevant fabrication procedures for manufacturing the shielding modules of the ITER Primary Wall. The fabrication of several Beryllium armored small scale mock-ups, reproducing representative portions of a Primary Wall panels, was also launched (Fusion Technol. (1998) 195). Further experimental activities were also programmed for investigating the thermal–mechanical behavior of these mock-ups at high heat flux and under thermal fatigue tests. In 2001, the ITER European Home Team decided to assign to ENEA a contract for the thermal fatigue testing of six mock-ups aiming at verifying the reliability of the Beryllium/Dispersion Strengthened Copper alloy/Stainless Steel and Beryllium/Precipitation hardened Copper alloy/Stainless Steel joints manufactured by solid Hot Isostatic Pressing (HIP) procedure (Technical Specification for the Thermal Fatigue Tests of Be protected EDA Mock-ups). The paper presents the results of the FEM thermal–mechanical analyses performed by ANSYS code and the progress of the first test campaign.

Detection of subcooled boiling heat transfer regimes up to critical heat flux by accelerometric equipment

Abstract The design of the plasma-facing components of the Next European Torus-International Thermonuclear Experimental Reactor fusion reactor has been addressing the highest requirement in the field of heat transfer thermal hydraulics. The more promising heat transfer technique, among those possible using water as coolant, is based on the subcooled boiling thermal hydraulics in the fully developed regime, with the highest heat transfer coefficient, but avoiding the reaching of the critical heat flux (CHF) and its consequent dangerous burn-out. To this aim an experimental activity was launched in order to optimize the material, the physical parameters and the structure geometry. Among others, an objective of this work is the development of an experimental system for the detection of the subcooled boiling phenomenon covering the whole heat transfer regime, on externally heated cylindrical channels, from the single phase up to the CHF. The principle is based on the recording, by using quartz accelerometers, of the bubble implosion noises transmitted by the structure in the proximity of the collapsing regions.

Experimental tests on Li-ceramic breeders for the helium cooled pebble bed (HCPB) blanket design

The Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) to be tested in ITER (International Thermonuclear Experimental Reactor) Reactor foresees the utilization of Lithiate ceramics as Tritium breeder in form of pebble beds. Since 1998, ENEA has launched many experimental activities for the evaluation of the breeder thermomechanics and the interaction between the pebble beds and the prismatic steel containment walls. Main objectives of these activities are the measurement of the pebble bed effective thermal conductivity, the wall heat transfer coefficient, the pressure loads and deformations on the lateral walls and their dependency from the mechanical constraints. The paper presents the progress of the second test campaign performed at ENEA Brasimone HE-FUS3 facility on Li4SiO4 pebbles.

Latest liquid lithium target design during the key element technology phase in the international fusion materials irradiation facility (IFMIF)

Abstract International Fusion Materials Irradiation Facility (IFMIF), being jointly developed by EU, JA, RF and US, is a deuteron–lithium (Li) stripping reaction neutron source for fusion materials testing. In 2002, a 3 year Key Element technology Phase (KEP) to reduce the key technology risk factors was completed. A liquid Li target has been designed to produce intense high energy neutrons (2 MW/m 2 ) up to 50 dpa/year by 10 MW of deuterium beam deposition which corresponds to an ultra high heat load of 1 GW/m 2 . This paper describes the latest design of the liquid Li target system reflecting the KEP results and future prospects.

Activities on IFMIF lithium target at ENEA

Abstract The status of R&D activity ongoing at ENEA on the lithium target system of the international fusion materials irradiation facility (IFMIF) is reported. The activity has been launched in year 2000 in the frame of IFMIF key element technology phase in order to reduce the key technology risk factors and to guarantee the required availability and reliability of the IFMIF liquid Li target system. The items discussed in the paper are related to the Li jet flow stability numerical analysis, water flow simulation experiment, backplate remote handling simulation and Li target safety analysis.

Tests on the integration of the ITER divertor dummy armour prototype on a simplified model of cassette body

Abstract In 1998, in the frame of the European R&D on ITER high heat flux components, the fabrication of a full scale ITER Divertor Outboard mock-up was launched. It comprised a Cassette Body, designed with some mechanical and hydraulic simplifications with respect to the reference body, and the actively cooled Dummy Armour Prototype (DAP). This DAP consists of the Vertical Target, the Wing and the Dump Target, manufactured by the European industry, which are integrated with the Gas Box Liner supplied by the Russian Federation Home Team. In order to simplify the manufacturing, the DAP was layered with an equivalent CuCrZr thickness simulating the real armour (CFC or W tiles). In parallel with the manufacturing activity, the ITER European HT decided to assign to ENEA the Task EU-DV1 for the “Component Integration and Thermal–Hydraulic Testing of the ITER Divertor Targets and Wing Dummy Prototypes and Cassette Body”.

Progress on fatigue characterization of ITER primary first wall mock-ups

Abstract In 2001, EFDA has assigned to ENEA a contract for the thermomechanical testing of six mock-ups of the ITER primary wall module. These small scale mock-ups, reproducing representative portions of the reference ITER primary wall panels, were fabricated during ITER EDA phase by solid hot isostatic pressing (HIPping) of an AISI 316L stainless steel back structure to a alumina dispersion strengthened (DS)-Cu alloy heat sink armored with beryllium tiles. The experimental program, carried-out at ENEA Brasimone CEF 1–2 thermal hydraulic facility, was focused on the thermal mechanical testing of these mock-ups aiming at verifying which tile geometry and manufacturing procedure assures the required reliability of the beryllium/DS-Cu alloy/SS joints at high incident heat flux (>0.8 MW/m 2 ) both at steady state and under thermal fatigue tests. The paper presents the progress in the experimental activity of the first test campaign and the main thermomechanical FEM analyses after the detachment and rupture of Beryllium tiles from one mock-up.

Experimental thermo-mechanical testing apparatus for a DEMO relevant blanket design of a ceramic cladded pin

Abstract The technological effort in supporting the development of a DEMO relevant Helium Cooled Solid Breeder Blanket (HCSBB) design, with an arrangement of the breeder inside the cladding tubes (BIT), foresees the qualification of some ceramic materials as tritium breeders. Some in-pile experiments are already carried out or are in progress, aiming at testing the materials as tritium release and selected mechanical properties. The facility allows to carry out some out-of-pile thermomechanical fatigue tests on a significant portion of one cladded ceramic pin. The breeder heat generation and the stresses are simulated by an internal electrical resistor. The radial thermal gradient is regulated by means of a proper air cooling flow with helium similitude. The internal hydraulics of the purge flow is fully reproduced, both chemically and thermally, and monitored in order to investigate the hot temperature corrosion effects on the sheath, its interaction with the ceramics and the pellet fractures.