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The paper presents a general overview of the Frascati Tokamak Upgrade (FTU) machine operation in the last two years. About 3500 pulses were performed (mostly in D/sub 2/) with a toroidal field between 2.5 and 8 Tesla and a plasma current between 0.35-1.3 MA, according to the experimental programmes. An availability analysis of the FTU plant was performed, which pointed out the number of pulses per day and the causes of delay. The main research activities concentrated on plasma MHD studies with different heating scenarios: lower hybrid waves current drive (LHCD) and electron cyclotron resonance heating (ECRH) were used both alone and together to investigate synergetic effects. Strong effort was devoted to high density (1/spl times/10/sup 20/ m/sup -3/) current drive experiments and investigating the possibility of obtaining and sustaining shear reversal configurations. Near term scientific and technological programmes are also reported.

After one year of shutdown devoted to the installation of the toroidal molybdenum limiter, the Frascati Tokamak Upgrade (FTU) has recently resumed operation. The experimental programme is focused on the three radiofrequency (RF) heating and non-inductive current drive systems: electron cyclotron resonant heating (ECRH), lower hybrid current drive (LHCD) and ion Bernstein wave heating (IBWH). The main aim of the experiments reported here is to demonstrate LHCD at high density and to sustain and heat reversed magnetic shear configurations, with a large reversal radius (r(s)/a approximate to 0.5). A transient heating of the shear reversed discharges achieved in ohmic conditions has been performed through ECRH. The results show that good confinement properties are maintained within the shear reversal surface. The FTU 8 GHz LH system was upgraded so that up to 1.4 MW of LH could be injected into the plasma. A value of 0.14 x 10(20) A . W-1 . m(-2) was obtained for the figure of merit of the CD efficiency. Furthermore, in ohmic and LH heated discharges a confinement time of up to 1.6 times the ITER89-P value was observed. The first shear reversed discharge driven by 0.8 MW of LHCD was produced at I-p = 350 kA, (n) over bar(e) = 7 x 10(19) m(-3); it shows a regime where regular m = 2, n = 1 relaxations with extremely large amplitude ((B) over tilde(theta)/B-theta = 2.5% at the plasma edge) coexist with a high temperature plasma (T-e0 = 1.6 keV).

Within the framework of the activities foreseen by the EUROfusion action on the cooling water activation assessment for a DEMO reactor equipped with a Water Cooled Lithium Lead Breeding Blanket (WCLL BB), the University of Palermo is involved in the investigation of the absorbed dose induced by the decay of nitrogen radioisotopes produced by water activation, in the main components (e.g. isolation valves) of both First Wall (FW) and Breeder Zone (BZ) cooling circuits. The aim of this work is to assess the spatial distribution of the absorbed dose in the DEMO Upper Pipe Chase (UPC), focusing the attention on the space neighbouring a typical isolation valve of the Primary Heat Transfer System (PHTS), taking into account a period of 7 full power years. To this end, a computational approach has been followed adopting MCNP5 Monte Carlo code. In particular, a totally heterogeneous neutronic model of a portion of the UPC has been set up, including the valve and the main FW and BZ PHTS piping, and the spatial distribution of nitrogen isotopes concentrations, previously assessed, have been used to model the photonic and neutronic sources. The results obtained, herewith presented and critically discussed, provided some information on the nuclear issues of the WCLL BB PHTS, to be considered as hints for the blanket design optimization.

In the frame of the activities promoted and encouraged by the EUROfusion Power Plant Physics and Technology (PPPT) department aimed at developing the EU-DEMO fusion reactor, strong emphasis has been recently addressed to the whole Balance of Plant (BoP) which represents the set of systems devoted to convert the plasma generated thermal power into electricity and to deliver it to the grid. Among these systems, the Divertor Primary Heat Transfer Systems (PHTSs) are intended to feed coolant to the two main components of the Divertor assembly, namely the Plasma Facing Components (PFCs) and the Cassette Body (CB). Since the DEMO Divertor must withstand high heat flux loads together with a considerable neutron deposited power, very tight tolerances may be allowed to the coolant inlet conditions. Therefore, the design of reliable PHTSs is of the utmost importance towards the development of an EU-DEMO fusion reactor. Within this framework, a study has been jointly carried out by University of Palermo, Ansaldo Nucleare and CREATE to design the Ex-Vessel PHTSs of both the PFCs and the CB for a DEMO reactor equipped with a Helium Cooled Pebble Bed Breeding Blanket. The paper describes criteria and rationale followed with the aim to achieve simple PHTS designs based on the adoption of easy-to-manufacture main components avoiding too much extrapolation from the state-of-art technology. Results of preliminary thermal-hydraulic calculations carried out to size heat exchangers, pressurizers, piping and pumps are presented and critically discussed, with particular attention to those integration, safety and feasibility constraints that may deeply affect the design of such components. Finally, the evaluation of PHTS key parameters as total pumping power and coolant inventory is reported.

EU-DEMO is a European project for the design and construction of the first nuclear reactor able to produce electrical energy thanks to thermonuclear fusion reactions. The electrical power system of this facility is particular demanding for several reasons. First, the electrical power required to operate DEMO is expected to be huge, especially in comparison to a conventional nuclear plant. Moreover, the operations are intrinsically intermittent, thus introducing specific problems in terms of energy balance and grid stability. This paper deals with a first preliminary study about the sizing and the design of the DEMO internal electrical power system, with a specific focus on the steady-state loads necessary to operate the plant. The design results are obtained and verified through a model for power flow analysis, implemented in DIgSILENT PowerFactory.

One of the generic designs of the nuclear fusion DEMO reactor proposed by the EUROfusion consortium foresees the development of a tritium breeding blanket (BB) relying on the use of the liquid-metal PbLi eutectic alloy as both neutron multiplier and tritium breeder, namely the water-cooled lithium lead (WCLL) BB, whose strengths and weaknesses are well known. This paper focuses the attention on one of the possible disadvantages of such a technology: the production of the highly radiotoxic radionuclide 210Po, which could become a safety issue to be accounted for. The 210Po concentration within the PbLi circuit has been assessed by solving a modified version of Bateman’s equations to consider the alloy circulation, so a one-dimensional convective fluid-dynamic model has been set up. Nuclear quantities have been evaluated by Monte Carlo neutron transport analyses using MCNP code and adopting a fully heterogeneous model of DEMO equipped with the WCLL BB. Moreover, rough sensitivity analyses have been performed to assess the influence on the results of the uncertainties related to the 209Bi radiative-capture cross section and the initial concentration of this nuclide which is present in the PbLi as an impurity. Results obtained have been critically discussed and some safety issues have been addressed to evaluate the possible hazard in case of a leak of PbLi accident.

The European DEMO conceptual design foresees four Breeding Blanket (BB) concepts that rely on different cooling and breeding technologies. As DEMO has been conceived to deliver net electricity to the grid, the choice of the blanket coolant plays a pivotal role in the reactor design having a strong influence on plant operation, safety and maintenance. Moreover, the machine pulsed operation makes the BB Primary Heat Transfer System (PHTS) the main hub of the DEMO Balance of Plant (BoP). Within this framework, a study has been carried out to design the Ex-Vessel PHTS of the Helium-Cooled Pebble Bed (HCPB) BB concept. The paper describes criteria and rationale followed with the aim to achieve a simple PHTS design based on the adoption of easy-to-manufacture main components. Results of preliminary thermal-hydraulic calculations carried out to size heat exchangers, piping and compressors are presented and critically discussed. Finally, the evaluation of PHTS key parameters as total pressure drops and total coolant inventory is reported to provide input relevant to the assessment of the impact of helium technology on the design integration and feasibility of DEMO BoP.

The EU-DEMO conceptual design is being conducted among research institutions and universities from 26 countries of European Union, Switzerland and Ukraine. Its mission is to realise electricity from nuclear fusion reaction by 2050. As DEMO has been conceived to deliver net electricity to the grid, the choice of the Breeding Blanket (BB) coolant plays a pivotal role in the reactor design having a strong influence on plant operation, safety and maintenance. In particular, due to the pulsed nature of the heat source, the Primary Heat Transfer System (PHTS) becomes a very important actor of the Balance of Plant (BoP) together with the Power Conversion System (PCS). Moreover, aiming to mitigate the potential negative impact of plasma pulsing on BoP equipment, for the DEMO plant is also being investigated a “heat transfer chain” option which envisages an Intermediate Heat Transfer System (IHTS) equipped with an Energy Storage System (ESS) between PHTS and PCS. Within this framework, a preliminary study has been carried out to analyse the thermal dynamic behaviour of the IHTS system for the Helium-Cooled Pebble Bed (HCPB) BB concept during pulse/dwell transition which should be still considered as the normal operating mode of a fusion power plant. Starting from preliminary thermal-hydraulic calculations made in order to size the main BoP components, the global performances of DEMO BoP have been quantitatively assessed focusing the attention on the attitude of the whole IHTS to smooth the sudden power variations which come from the plasma. The paper describes criteria and rationale followed to develop a numerical model which manages to simulate simple transient scenarios of DEMO BoP. Results of numerical simulations are presented and critically discussed in order to point out the main issues that DEMO BoP has to overcome to achieve a viable electricity power output.