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Within the framework of the research and development activities supported by the ITER Organization on the blanket system issues, an intense analysis campaign has been performed at the University of Palermo with the aim to investigate the thermal-hydraulic behaviour of the cooling system of a standard 20° sector of ITER blanket during the draining transient operational procedure. The analysis has been carried out following a theoretical-computational approach based on the finite volume method and adopting the RELAP5 system code. In a first phase, attention has been focused on the development and validation of the finite volume models of the cooling circuits of the most demanding modules belonging to the standard blanket sector. In later phase, attention has been put to the numerical simulation of the thermal-hydraulic transient behaviour of each cooling circuit during the draining operational procedure. The draining procedure efficiency has been assessed in terms of both transient duration and residual amount of coolant inside the circuit, observing that the former ranges typically between 40 and 120 s and the latter reaches at most ∼8 kg, in the case of the cooling circuit of twinned modules #6–7. Potential variations to operational parameters and/or to circuit lay-out have been proposed and investigated to optimize the circuit draining performances. In this paper, the set-up of the finite volume models is briefly described and the key results are summarized and critically discussed. [ABSTRACT FROM AUTHOR]

The blanket system is the ITER reactor component devoted to providing a physical boundary for plasma transients and contributing to thermal and nuclear shielding of vacuum vessel, magnets and external components. It is expected to be subjected to significant heat loads under nominal conditions and its cooling system has to ensure an adequate cooling, preventing any risk of critical heat flux occurrence while complying with pressure drop limits. At the University of Palermo a study has been performed, in cooperation with the ITER Organization, to investigate the steady state hydraulic behaviour of the ITER blanket standard sector cooling system. A theoretical–computational approach based on the finite volume method has been followed, adopting the RELAP5 system code. Finite volume models of the most critical blanket cooling circuits have been set-up, realistically simulating the coolant flow domain. The steady state hydraulic behaviour of each cooling circuit has been investigated, determining its hydraulic characteristic function and assessing the spatial distribution of coolant mass flow rates, velocities and pressure drops under reference nominal conditions. Results obtained have indicated that the investigated cooling circuits are able to provide an effective cooling to blanket modules, generally meeting ITER requirements in term of pressure drop and velocity distribution, except for a couple of circuits that are being revised. [ABSTRACT FROM AUTHOR]

This paper describes the main activities carried out in the frame of EU-DA prequalification for the supply of Normal Heat Flux (NHF) First Wall (FW) panels to ITER. A key part of these activities is the manufacturing development, the fabrication and the factory acceptance tests of a reduced scale FW prototype (Semi-Prototype (SP)) of the NHF design. The SP has a dimension of 221 mm × 665 mm, corresponding to about 1/6 of a full-scale panel, with six full-scale “fingers” and bearing a total of 84 beryllium tiles. It has been manufactured by the AREVA Company in France. The manufacturing process has made extensive use of Hot Isostatic Pressing, which was developed over more than a decade during the ITER Engineering Design Activity phase. The main manufacturing steps for the Semi-Prototype are recalled, with a summary of the lessons learned and the implications with regard to the design and manufacturing of the full-scale prototype and of the series fabrication of the EU-DA share of the ITER first wall (215 NHF panels). The fabricated SP is then tested under High Heat Flux (HHF) in the dedicated test facility of JUDITH-II in Forschungszentrum Jülich, Germany. The objective of the HHF testing is the demonstration of achieving the requested performance under thermal fatigue. The test protocol and facility qualification are presented and the behaviour of the fingers under the 7500 cycles at 2 MW/m 2 is described in detail. [ABSTRACT FROM AUTHOR]