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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.

The target system is one of the key issues for the facilities aimed at the production of neutron-rich radioactive ion beams. In the framework of the SPES project (Study for the Production of Exotic Species), the possibility of using a target configuration with a proton beam (40MeV, 0.2mA) directly impinging on multiple uranium carbide disks is investigated. The 238U fission fragments constitute the source for the exotic beams and for this purpose the disks are placed inside a graphite box at 2000 °C. The target is split into several thin disks in order to allow the cooling of the system by thermal radiation. In this way about ∼ 1013 fissions s -1 are obtained with a relative simple system and with relative low costs. Further steps have been performed compared to previous publications and now all the main parameters of the system have been analysed by means of calculation codes: the fission rates and the fission fragment distribution; the power deposition and the thermal analysis; the thermo-mechanical behaviours of the disks; the effusive and diffusive extraction release properties of the target.

During the three year key element technology phase of the International Fusion Materials Irradiation Facility (IFMIF) project, completed at the end of 2002, key technologies have been validated. In this paper, these results are summarized. A water jet experiment simulating Li flow validated stable flow up to 20 m/s with a double reducer nozzle. In addition, a small Li loop experiment validated stable Li flow up to 14 m/s. To control the nitrogen content in Li below 10 wppm will require surface area of a V–Ti alloy getter of 135 m 2 . Conceptual designs of diagnostics have been carried out. Moreover, the concept of a remote handling system to replace the back wall based on `cut and reweld' and `bayonet' options has been established. Analysis by FMEA showed safe operation of the target system. Recent activities in the transition phase, started in 2003, and plan for the next phase are also described.

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.

This paper describes the status of the liquid lithium (Li) target facility in the key element technology phase (KEP) of the International Fusion Materials Irradiation Facility (IFMIF). The IFMIF is being jointly developed by the 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 reactors. A 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 a Li simulation experiment by water jet, Li flow experiment, corrosion/erosion, remote handling of the target assembly, and safety analysis. In addition to the KEP tasks, detailed design of the target is being performed.

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.

In the IFMIF Lithium target system, the total lithium inventory is about 21 m 3 and the impurities in the lithium are controlled by a cold trap and a hot trap. From the safety standpoint, the control of tritium and beryllium-7 concentration in the lithium is the most important issue. Both elements are to be controlled by a cold trap. As an option for the tritium, a hot trap is also considered. Other impurities such as oxygen, nitrogen and carbon which will affect the compatibility of lithium and loop materials are also controlled by the cold and hot traps to an acceptable level. A swamping method to control the tritium concentration is analyzed.

The lithium target facility design has been developed in the Conceptual Design Activities for the International Fusion Materials Irradiation Facility (IFMIF). As for the lithium target design, the replaceable backwall concept is selected as a reference design with the cell environment of 10{sup -1}Pa, and the free jet and Fusion Materials Irradiation Test Facility (FMIT) type design are carried as the option. In the lithium loop design the single loop is layed out to provide the lithium flow to the two targets with redundant components including tanks, heat exchangers, lithium purification systems, and lithium monitoring systems for high availability. Total lithium flow capability of 120 liter/s is provided by single electromagnetic pump, and the total lithium inventory is 22,300 liters. 9 refs., 5 figs., 2 tabs.

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 blanket option featured by helium cooling and lithiated ceramics inside tubes is under development at CEA/ENEA within the framework of the European Test Blanket Program. A first conceptual design...

In the framework of the SPES project at LNL [A. Bracco, A. Pisent (Ed.), REP 181/02, LNL-INFN, 2002], the realization of a direct ISOL Target for a mid-term radioactive ion beam facility is in progress. Using a primary proton beam of energy 40 MeV and intensity 0.2 mA, a high number of fission products will be obtained in the SPES multi-foil uranium carbide target, keeping a low power density deposition in the refractory matrix [A. Andrighetto, S. Cevolani, C. Petrovich, Eur. Phys J. A 25 (2005) 41]. The exotic species produced by Uranium fission in the target are collected in the ion source after the diffusion and the effusion processes. When short lived isotopes are produced it is very important to optimize the release properties of the target. To this purpose the RIBO code (radioactive ion beam optimiser) [M. Santana Leitner, A Monte Carlo Code to Optimize the Production of Radioactive Ion Beams by the ISOL Technique, PhD. Thesis, UPC-ETSEIB/CERN] has been used in order to estimate the target release efficiency for some neutron-rich nuclei. A SiC prototype of the target was recently produced at LNL and tested at ORNL using a 42 MeV proton beam. The yield of some aluminum isotopes was measured as a function of the target temperature. Some preliminary results of the data analysis will be presented.

A solution for a target aimed at the production of exotic nuclei will be analyzed. In the proposed configuration a primary proton beam (40 MeV, 0.2 mA) directly impinges on uranium carbide disks. The exotic nuclei are produced as a result of high energy fissions in 238 U compounds. The system has been conceived in order to obtain both a high number of neutron-rich atoms, originating from about 10 13 fissions s −1 and a low power deposition in the target. The target container is held at about 2000 °C in order to maximize the number of extracted fission fragments.

A R&D study for the realization of a Direct Target is in progress within the SPES project for RIBs production at the Laboratori Nazionali of Legnaro. A proton beam (40 MeV energy, 0.2 mA current) is supposed to impinge directly on a UCx multiple thin disks target, the power released by the proton beam is dissipated mainly through irradiation. A SiC target prototype with a 1:5 scale has been developed and tested. Thermal, mechanical and release calculations have been performed to fully characterize the prototype. An online test has been performed at the HRIBF facility of the Oak Ridge National Laboratory (ORNL), showing that our SiC target can sustain a proton beam current considerably higher than the maximum beam current used with the standard HRIBF target configuration.

The construction of the Radioactive Ion Beam (RIB) Facility SPES‐DT, within the framework of the new European RIB panorama is under study at the Laboratori Nazionali di Legnaro. The present project will be devoted to the production of neutron rich exotic nuclei, by using the fission process induced by a 40 MeV proton beam, 200 μA current, on a multi‐sliced Uranium Carbide (UCx) target. The UCx multiple thin disks target, developed at LNL is designed so that the power released by the proton beam is mainly dissipated by irradiation. Following the idea of the existing HRIBF facility at Oak Ridge National Laboratory (USA) where a proton primary beam of 40 MeV is also used, our target configuration is an evolution which permits to sustain a higher power on target. A high number of fission products (up to 1013 fission/s) will be obtained, still keeping a low power density deposition inside the target. The whole facility, together with the details on the Direct Target configuration, will be described. Thermo‐mec...

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 proposed configuration is constituted by a primary proton beam (40 MeV, 0.2 mA) directly impinging on uranium carbide disks inserted within a cylindrical carbon box. This system has been conceived to obtain both a high number of neutron rich atoms (originated from about 1013 fissions/s) and a low power deposition in the target. In order to extract the fission fragments, the box has to be hold at 2000○C. The thermal analysis shows the capability of the thermal radiation to cool the disks with a reasonable margin below the material melting point. Moreover, the analyses of the thermo-mechanical behaviour and of the effusion times confirm the promising features of this target configuration.

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.

An accelerator-based source of thermal neutrons for treating cutaneous melanoma is foreseen in the framework of the SPES project,1 proposing a facility for exotic nuclei production at the Laboratori Nazionali di Legnaro (LNL) in Italy. The goal of the project, scheduled for the next five years (1999–2003), is the construction of a l0MeV proton/deuteron accelerator capable of delivering a current of 30 mA, consisting of a 5MeV Radio-Frequency Quadrupole (RFQ) and of a Drift-Tube Linac (DTL). Part of the beam (1-3mA) delivered by the RFQ to the linear accelerator will be used for neutron production for BNCT, via the 9Be(p,n)9B and/or the 9Be(d,n)10B reactions. The BNCT facility which will be built at the LNL is intended to be in the framework of the TERA program for the realization of hadrontherapy facilities in Italy.2

A Direct Target for a mid-term RIB ISOL-type facility is being developed at LNL, in the framework of the R&D for the SPES project [1]. Using a 40 MeV proton beam impinging on a UCx thick target of 2.5 g/cm3 density, a production rate of 1013 in target fissions per second is expected [2]. The crucial point, when short-lived isotopes are produced in the target, is to build systems (target + ion source) with good release properties and high efficiency. Monte Carlo simulations were performed using the GEANT4 toolkit [3] and the RIBO code [4] in order to optimize our target geometry and to estimate the average release time.

The European ceramic breeder-inside-tube (B.I.T.) blanket module design proposed for the DEMO plant is composed of a poloidal tubular vessel with 19-pin bundle of cladded rods equipped with a spacer system surrounded by beryllium blocks. In 1990 CEA and ENEA initiated a collaboration on a common conceptual design and test activities on this blanket option. In the frame of the European Test Blanket Programme this concept is candidate for the selection, scheduled by the end of 1994, of the two that will be tested in the NET/ITER reactor. The non nuclear out of pile tests are part of the whole experimental activity planned in order to investigate some critical design issues.

The advanced ENEA poloidal solid breeder-in-tube (BIT) blanket design for a DEMO fusion reactor is presented. Its main operating conditions and features are max. neutron fluence = 3×1022 n/cm2; max. beryllium vol. swelling = 4%; max. beryllium differential swelling = 3%; module poloidal length = 8.5 m; helium inlet/outlet temperature = 250/500°C; pumping power percentage >4%; max. structure temperature >550°C; ceramic breeder min/max. temperature 450/750°C; purge circuit separated from the coolant circuit; coaxial manifolding system; 3-D value of the t.b. r ≈ 1.20; and thermal cycle efficiency 35%.