The European DEMO is considered to be the nearest-term fusion reactor with the aim to generate several hundred MWs of net electricity, operate with a closed tritium fuel-cycle (achieving the tritium self-sufficiency) and qualify technological solutions for a Fusion Power Plant. Two Breeding Blanket (BB) concepts relying on different cooling and breeding technologies are considered for the baseline design: the Helium Cooled Pebble Bed BB and the Water Cooled Lithium Lead BB. The selection of the BB type is a key factor for the development of the whole DEMO plant design and, in particular, of those systems having the responsibility to remove the plasma generated thermal power and its conversion in mechanical and finally electrical energy, namely: the Primary Heat Transport System (PHTS), the Power Conversion System (PCS) and the Intermediate Heat Transport System (IHTS) - equipped by an Energy Storage System (ESS) - in between PHTS and PCS, which is introduced with the purpose of buffering some energy produced by the pulsed plasma source to allow a continuous production of electricity. This work deals with the preliminary thermo-mechanical design of the once through molten salt/water steam generator exhausting the thermal power during pulse operation and of the helium/molten salt intermediate heat exchanger of the IHTS. The main challenges in the design are given by the implementation of the design rules according to nuclear standards, the cyclic operating load and, for what the main HX is concerned, the novelty of the application as well as the large dimensions of parts. The design criteria and the main design analyses developed to achieve the best simple but robust design are discussed.

One of the main challenges, within the European Fusion Roadmap, in view of the construction of a demonstration plant (DEMO, the first nuclear fusion power plant able to provide power to the electricity grid around 2050), is the thermal power on the divertor. ITER plans to test the actual possibilities of a "standard" divertor operating in a plasma fully detached condition, i.e. no contact between plasma and first wall of the vessel. This solution could be unsuitable to be extrapolated to the operating conditions of DEMO and future reactors; then the problem of thermal loads on the divertor may remain particularly critical in the road to the realization of the reactor. For this reason, a specific project has been launched, aimed to define and design a Tokamak named "DTT (Divertor Tokamak Test)". This Tokamak has to carry out a number of scaled experiments, to be integrated with the specific physical condition expected and technological solutions included in DEMO. DTT should retain the possibility of testing different divertor magnetic configurations, including liquid metal divertor targets, and other possible solutions promising to face with the power exhaust problem. The construction has recently been approved by the Italian government. DTT will be a high field superconducting toroidal device (6 T) carrying plasma current up to 5.5 MA in pulses with length up to about 100 s and with 45 MW of additional heating power. The nominal cross section is elongated with a major radius R=2.11 m and a minor radius a=0.64m. The DTT parameters are selected so as to reproduce edge conditions as close as possible to those expected in DEMO (in terms of a set of dimensionless parameters characterizing the physics of Scrape Off Layer, SOL, and of the divertor region), while fully fitting (again, in terms of the dimensionless parameters) with DEMO bulk plasma performance. Maximum flexibility is guaranteed, within the limits of a given budget and a tight time schedule consistent with the needs of the European Road Map. This paper describes the status of the design activities of DTT. Emphasis is given on an integrated design approach, illustrate the rationale for the design choices, focusing on the main components, namely magnet system, plasma scenarios, vacuum vessel, in-vessel components, thermal shield, neutron shield, and additional heating system.

One of the main challenges, within the European Fusion Roadmap, in view of the construction of a demonstration plant (DEMO, the first nuclear fusion power plant able to provide power to the electricity grid around 2050), is the thermal power on the divertor. ITER plans to test the actual possibilities of a "standard" divertor operating in a plasma fully detached condition, i.e. no contact between plasma and first wall of the vessel. This solution could be unsuitable to be extrapolated to the operating conditions of DEMO and future reactors; then the problem of thermal loads on the divertor may remain particularly critical in the road to the realization of the reactor. For this reason, a specific project has been launched, aimed to define and design a Tokamak named "DTT (Divertor Tokamak Test)". This Tokamak has to carry out a number of scaled experiments, to be integrated with the specific physical condition expected and technological solutions included in DEMO. DTT should retain the possibility of testing different divertor magnetic configurations, including liquid metal divertor targets, and other possible solutions promising to face with the power exhaust problem. The construction has recently been approved by the Italian government. DTT will be a high field superconducting toroidal device (6 T) carrying plasma current up to 5.5 MA in pulses with length up to about 100 s and with 45 MW of additional heating power. The nominal cross section is elongated with a major radius R=2.11 m and a minor radius a=0.64m. The DTT parameters are selected so as to reproduce edge conditions as close as possible to those expected in DEMO (in terms of a set of dimensionless parameters characterizing the physics of Scrape Off Layer, SOL, and of the divertor region), while fully fitting (again, in terms of the dimensionless parameters) with DEMO bulk plasma performance. Maximum flexibility is guaranteed, within the limits of a given budget and a tight time schedule consistent with the needs of the European Road Map. This paper describes the status of the design activities of DTT. Emphasis is given on an integrated design approach, illustrate the rationale for the design choices, focusing on the main components, namely magnet system, plasma scenarios, vacuum vessel, in-vessel components, thermal shield, neutron shield, and additional heating system.

Fusion-fission hybrid reactors (FFHR), which consist of a neutron-producing fusion core surrounded by a fission blanket, keep a sustained interest because of their potential to address energy production before the availability of pure fusion reactors and also because of their capability of fuel supply and waste management. The U.S. Department of Energy (DOE) sponsored a workshop in 2009 in order to assess the potentiality of fusion-fission reactors and, among the others alternative to the Tokamak as fusion core, the Reversed Field Pinch (RFP) was considered, even if not on the same timeline of the Tokamak. On the other hand, the interest in the RFP as fusion core resides, inter alia, in its possibility to reach ignition in a pure ohmic way with a "light" toroidal field winding, which simplifies significantly the hybrid reactor layout and operation and strongly decreases costs with respect to other fusion core solutions. New studies are now in progress in order to revisit the status and the potentiality of the RFP as fusion core in FFHR, taking into account (i) the recent progress in RFP physics brought by the results of the RFX-mod experiment (R=2, a=0.46, Ip=2MA), mainly about MHD modes control, (ii) the expected performance improvements resulting from the ongoing upgrade of the machine. Starting from RFX-mod results and scaling laws, and considering the estimated improvements of the performances derived from the machine enhancements underway, the possibility to realize a preliminary pilot experiment in which a RFP is the neutron fusion source is analysed. The research needs for this advanced solution are updated with respect to the 2009 workshop.

Fusion-fission hybrid reactors (FFHR), which consist of a neutron-producing fusion core surrounded by a fission blanket, keep a sustained interest because of their potential to address energy production before the availability of pure fusion reactors and also because of their capability of fuel supply and waste management. The U.S. Department of Energy (DOE) sponsored a workshop in 2009 in order to assess the potentiality of fusion-fission reactors and, among the others alternative to the Tokamak as fusion core, the Reversed Field Pinch (RFP) was considered, even if not on the same timeline of the Tokamak. On the other hand, the interest in the RFP as fusion core resides, inter alia, in its possibility to reach ignition in a pure ohmic way with a "light" toroidal field winding, which simplifies significantly the hybrid reactor layout and operation and strongly decreases costs with respect to other fusion core solutions. New studies are now in progress in order to revisit the status and the potentiality of the RFP as fusion core in FFHR, taking into account (i) the recent progress in RFP physics brought by the results of the RFX-mod experiment (R=2, a=0.46, Ip=2MA), mainly about MHD modes control, (ii) the expected performance improvements resulting from the ongoing upgrade of the machine. Starting from RFX-mod results and scaling laws, and considering the estimated improvements of the performances derived from the machine enhancements underway, the possibility to realize a preliminary pilot experiment in which a RFP is the neutron fusion source is analysed. The research needs for this advanced solution are updated with respect to the 2009 workshop.