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This paper describes the status of liquid lithium (Li) target facility in Key Element technology Phase (KEP) of International Fusion Materials Irradiation Facility (IFMIF). The IFMIF is being jointly developed by European Union (EU), Japan (JA), Russian Federation (RF) and US to provide an accelerator-based D-Li neutron source for testing the candidate materials for fusion reactor. Key issue of the Li target is to obtain a stable liquid Li flow with a speed of 20 m/s under a deuterium beam deposition of 10 MW. In the KEP, 19 tasks for the Li target are proposed and shared by EU, JA and RF. These tasks are Li flow experiment, Li simulation experiment by water jet, corrosion/erosion, remote handling, and safety analysis. In addition to the KEP tasks, detailed design of the target is being performed on transient thermal analysis of the Li , remote handling of the target assembly and safety analysis of the Li loop.

This paper presents an assessment of ATHENA/Modl, an extension of the RELAP5 code able to represent a number of cooling fluids, on experimental data provided by the helium facility HE_FUS3. The facility was selected to perform experimental campaigns on thermomechanics and thermal-hydrau]ics within the frame of the DEMO HCPB Blanket design programme. As a first step the characterisation tests of the facility have been used to develop an input model for ATHENA. Special models have been adopted in order to simulate some prototypical components for which the generic component models available in the code are inadequate. Later on, a preliminary assessment of this model has been performed, both for steady state and transient conditions, on the first tests conducted in the facility. In spite of the numerical weaknesses detected during the analysis, which have required special caution in the definition of the boundary condetions, the stabilised steady state has been achieved in good agreoment with the experimental data, and two accident transients of LOFA have been qualitatively well reproduced.

Within the European Fusion Technology Programme, for the period 1997-1998, significant R&D activities related to the DEMO Reactor Helium Cooled Pebble Bed Blanket (HCPB) concept are being carried-out. The conceptual design and layout of the Test Blanket Module (TBM) to be tested in the ITER the fabrication of the related test sections and the Out-of-Pile testing activities are performed in the Work Package WP B2.1 and WP B3.2. The reference facility for testing the TBM medium scale mock-ups is the HE-FUS3 facility at ENEA Brasimone. These experiments will have the main objectives to investigate the thermal and mechanical behaviour of the TBM module under steady state, cyclic power and conditions. The paper presents the TBM mock-up design assumptions the preliminary calculations, the envisaged manufacturing procedures and a draft programme for the fabrication and testing of the mock-ups.

Since the 1997 ENEA has launched two out-of-pile experimental activities related to the design of the EU-HT Breeding Blanket (BB) for ITER and the design of the Helium Cooled Pebble Bed Blanket (HCPB) and its Test Blanket Module (TMB) (to be tested in ITER) for DEMO. Both Breeding Blanket designs are characterised for having adopted similar Beryllium binary pebble bed arrangement as reference neutron multiplier material. On the contrary the breeder material and its arrangement in the Blanket module is different.The ENEA out-of-pile experiments have the main objective to investigate the thermo-mechanical behaviour of the blanket modules aiming at obtain the experimental confirmation that the basic lay-out of the module accommodates the Internal heat loads at the reference geometrical and thermal hydraulic boundary conditions.The tests will be carried-out at ENEA, Brasimone Labs in the THESIS apparatus of the CEF1-2 thermal-hydraulic facility and on the HE-FUS3 thermal-hydraulic facility.The paper presents both the designs of the experiments and the related mock-ups.The first thermal-hydraulic and thermo-mechanical calculations with a tentative to adopt a model to simulate pebble beds in a FEM code are also presented.

The Tokamak Cooling Water System of nuclear facility has the function to remove heat from plasma facing components maintaining coolant temperatures, pressures and flow rates as required and, depending on thermal-hydraulic requirements, its systems are defined as High Energy Piping (HEP) because they contain fluids, such as water or steam, at a pressure greater than or equal to 2.0 MPa and/or at a temperature greater than or equal to 100 °C, or even gas at pressure above the atmospheric one. The French standards contemplate the need to consider the whipping effect on HEP design. This effect happens when, after a double ended guillotine break, the reaction force could create a displacement of the piping which might affect adjacent components. A research campaign has been performed, in cooperation by ITER Organization and University of Palermo, to outline the procedure to check whether whipping effect might occur and assess its potential damage effects so to allow their mitigation. This procedure is based on the guidelines issued by U.S. Nuclear Regulatory Commission. The proposed procedure has been applied to the analysis of the whipping effect of divertor primary heat transfer system HEP, using a theoretical–computational approach based on the finite element method.

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.

The ITER Organization (IO) developed a thermal-hydraulic (TH) model of the complex first wall and blanket (FW/BLK) cooling channels to determine gas flow rate and pressure required to effectively blow out the water in the FW/BLK. In addition, US ITER conducted experiments for selected geometries of FW/BLK flow channels to predict the blowout parameters. The analysis indicates that as low as 2 MPa of pressure difference over the blanket modules will ensure substantial evacuation of the water in blankets with just a few percent remaining in the blanket flow channels. A limited validation study indicates that the analysis yields less conservative results to compare against data collected from experiments. Therefore, the designed blow out flow of the drying system was selected with a large margin above the measured values to ensure the blow out operation.

The ITER diagnostic Upper Port Plug (UPP) is a water-cooled stainless steel structure aimed to integrate within vacuum vessel the plasma diagnostic systems, shielding them from neutron and photon irradiation. Due to the very intense heat loads expected, a proper cooling circuit has been designed to ensure an adequate UPP cooling with an acceptable thermal rise and an unduly high pumping power and to perform its draining and drying procedure by injection of pressurized nitrogen. A theoretical research activity has been launched at the Department of Nuclear Engineering of the University of Palermo aiming to investigate the hydraulic behaviour of the UPP Trapezoid Section cooling circuit under steady state conditions and during its draining and drying transient procedure. The research activity has been performed following a theoretical–computational approach and adopting the RELAP5 thermal–hydraulic system code. The Trapezoid Section cooling circuit characteristic functions have been derived under steady state conditions at various coolant temperatures for both the coolant flow paths at the present under consideration for this circuit. The distributions of coolant mass flow rates along the channels of the cooling circuit have been calculated too. Results show that the flow path characterized by right plate inlet has improved hydraulic performances. The transient behaviour of the Trapezoid Section cooling circuit has been investigated during the draining and drying operational transient procedure, considering realistic operative scenarios, for both the coolant flow paths at the present under consideration for the cooling circuit. In particular, it has been found out that the recently proposed flow path seems to allow the complete draining of the Trapezoid Section circuit, eliminating the need for the drying procedure.

Within the framework of the R&D activities promoted by European Fusion Development Agreement on the Helium-Cooled Pebble Bed Test Blanket Module to be irradiated in ITER, ENEA Brasimone and the Department of Nuclear Engineering of the University of Palermo (DIN) performed intense research activities on the modelling of the thermo-mechanical behaviour of both beryllium and lithiated ceramics pebble beds, that are envisaged to be used, respectively, as neutron multiplier and tritium breeder. In particular, the DIN developed a thermo-mechanical constitutive model for these pebble beds to be validated against the HEXCALIBER mock-up test campaign, carried out at the ENEA HE-FUS3 facility. The paper presents the main results of the mock-up experimental tests and of their numerical simulations performed adopting the finite element method, which allowed the DIN constitutive model to be assessed and validated.

In the frame of the activities related to ITER divertor R&D, ENEA CR Brasimone was in charge by EFDA (European Fusion Development Agreement) to investigate the thermal-hydraulic behaviour of the full-scale divertor plasma facing components, i.e. the outer vertical target, the inner vertical target and the dome-liner, both in steady state and during draining and drying transient. The investigation was performed by means of both experimental test campaigns performed at ENEA CR Brasimone and theoretical simulation developed in RELAP5 Mod.3.3 environment at the Department of Nuclear Engineering of the University of Palermo (DIN). This paper presents the achieved experimental results for both steady state and transient tests. Moreover, the level of the implementation of the predictive hydraulic model, based on RELAP5 code, as well as its results are described, discussed and compared with the experimental ones.

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.

One of the most challenging components of ITER is the divertor devoted at controlling the characteristics of the plasma boundary, exhausting the α particles and reducing the impurities in the plasma. The thermal–hydraulic design of the divertor is particularly, demanding because of the high heat loads and the cooling flow margin in the plasma-facing components (PFCs). The pressure drop is limited by the pumping power and also avoiding the risk of reaching critical heat flux (CHF). Furthermore, for maintenance operation foreseen, each single divertor cassette should be drained and dried before withdrawing it out from the vacuum vessel. To address these requirements, European Fusion Design Agreement (EFDA) has launched a research programme aimed at investigating the theoretical thermal–hydraulic behaviour of the whole divertor cassette, both in the operation and during the water discharge. The study has been carried out with the RELAP5 computer code and the results obtained are herewith presented.

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.