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This paper describes the latest design of liquid lithium (Li) target system in International Fusion Materials Irradiation Facility (IFMIF). IFMIF is an accelerator-driven intense neutron source for fusion reactor materials testing. Two 40 MeV deuteron beams with a total current of 250 mA strikes a liquid Li target circulating in a Li loop. The Li target is to provide a stable Li jet at a speed of 10–20 m/s to handle an averaged heat flux of 1 GW/m 2 . A concaved Li flow is applied to avoid Li boiling. A cold trap and two kinds of hot trap are applied to control impurities (T, 7 Be, C, O, N) below permissible levels. A back wall made of RAFM is located in the most severe region of neutron irradiation (50 dpa/year). System availability requires more than 95% during plant lifetime. EVEDA tasks including Li loop are now being performed under the Broader Approach.

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/m2) up to 50 dpa/year by 10 MW of deuterium beam deposition which corresponds to an ultra high heat load of 1 GW/m2. This paper describes the latest design of the liquid Li target system reflecting the KEP results and future prospects.

The Conceptual Design Activity (CDA) for the International Fusion Materials Irradiation Facility (IFMIF) has been conducted. For the IFMIF Target Facility, the conceptual designs of the following two main components have been performed. The design concept of IFMIF utilizes a high energy deuteron beam of 30–40 MeV and total current of 250 mA, impinging on a flowing lithium jet to produce high energy neutrons for irradiation of candidate fusion materials. (1) The target assembly: The kinetic energy of the deuteron beam is deposited on a Li-jet target and neutrons are produced through the d–Li stripping reaction in this target. The assembly is designed to get a stable lithium jet and to prevent the onset of lithium boiling. For 40-MeV deuteron beam (total current of 250 mA) and a beam footprint of 5 × 20 cm2 lithium jet dimensions are designed to be 2.5 cm thick and 26 cm wide. The lithium jet parameters are given. (2) Lithium loop: The loop circulates the lithium to and from the target assembly and removes the heat deposited by the deuteron beam containing systems for maintaining the high purity of the lithium required for radiological safety and to minimize corrosion. The maximum lithium flow rate is 130 l/s and the total lithium inventory is about 21 m3. The IFMIF policy requires that the lithium loop system be designed to guarantee no combustion of lithium in the event of a lithium leak. This can be achieved by use of multiple confinement of the lithium carrying components. The radioactive waste generated by the Target Facilities is estimated.

The analyses performed on the rod bundles of the CEA/ENEA poloidal Breeder In Tube (BIT) blanket have shown the necessity to investigate experimentally the helium coolant thermal-hydraulic behaviour. The paper describes the design of the experiment splitted into test facility and test section design. In the frame of the latter item, a method has been developed in order to reduce the geometrical independent variables of the bundle. The method is based on the minimization of the bundle transversal thermal gradient. If adopted as design rule, it will allow the minimization of the bundle hot-spot factor.

The European ceramic breeder-inside-tube blanket design proposed for the DEMO plant is featured by helium-cooled poloidal breeder modules consisting of a shrouded bundle of breeder tubes contained in a pressure tube and surrounded by beryllium blocks. This kind of breeder modules raises some problems as far as the control of pressure losses and hot spots is concerned, problems which might, a priori, jeopardize the feasibility and performance of the blanket. Results of first thermal-hydraulic analyses, performed by means of heat transport and subchannel codes, prove however to be quite encouraging in this respect. The elimination of remaining uncertainties — not crucial for the viability of the proposed module design — will nevertheless require specific experiments.

The DEMO specifications defined so as to ensure the consistency of the various blanket conceptual design studies performed within the framework of the European Test Blanket Program me have been recently updated. A very first attempt has been made to adapt the European Ceramic Breeder Inside-Tube DEMO blanket to these new specifications. Two solutions have been investigated. The first would ensure tritium self-sufficiency of the plant with a large safety margin. The other one, which fully preserves the design simplicity and reliability of the initial design, appears to be somewhat marginal from the tritium breeding capability point of view, but to offer good improvement prospects.

A possible solution for a target system aimed at the production of exotic nuclei as a result of high-energy fissions in 238U compounds has been analyzed. The configuration proposed is constituted by a primary proton beam (40 MeV, 0.4 mA) directly impinging on a UC2 multiple-disc target inserted within a cylindrical tungsten box. In order to extract the fission fragments, the tungsten box has to be kept at 2000 °C. This system has been conceived to obtain both a high number of fission fragments (about 2 . 1013 atoms/s) and a quite low power deposition in the target. The power release and the fragment distribution have been calculated by means of the Monte Carlo code MCNPX. The thermal analysis of the proposed configuration shows the capability of the thermal radiation to cool the discs with a reasonable margin below the material melting point. Moreover, the possibility of increasing such margin with simple modifications of the target design is shown by means of parametric analyses. The thermal analysis of the tungsten box, also cooled by radiation, points out the necessity to heat it and/or shield it thermally, in order to take it at the requested temperature. Preliminary calculations of the target-induced activity have also been performed.

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.