M. Gasparotto

Overview of Materials Research for Fusion Reactors

Materials research for fusion reactors is overviewed from Japanese, EU and US perspectives. Emphasis is placed on programs and strategies for developing blanket structural materials, and recent highlights in research and development for reduced activation ferritic martensitic steels, vanadium alloys and SiC/SiC composites, and in mechanistic experimental and modeling studies. The common critical issue for the candidate materials is the effect of irradiation with helium production. For the qualification of materials up to the full lifetime of a DEMO and Power Plant reactors, an intense neutron source with relevant fusion neutron spectra is crucial. Elaborate use of the presently available irradiation devices will facilitate efficient and sound materials development within the required time scale.

Transition From Construction to Operation Phase of the Wendelstein 7-X Stellarator

Assembly of the superconducting stellarator Wendelstein 7-X is well advanced, and commissioning of the device is being prepared. A first draft of the commissioning tasks has been developed and will be discussed in this paper.

Overview of fusion nuclear technology in Europe

Abstract The fusion nuclear technology programme in the EU is focussed on materials and breeding blankets development, tritium and high heat flux component technologies. A strong effort is also devoted to the validation of the design of an intense 14 MeV neutron source (IFMIF). The material programme includes the development of reduced activation ferritic martensitic steel (EUROFER) to be used as structural material in a DEMO reactor, and potentially more attractive higher performance materials: ODS and SiC/SiC composites. The breeding blanket activities are focussed in the preparation of the two European Test Blanket Moduli to be installed in ITER. The Fuel Cycle activities for ITER include development of the torus exhaust cryopump, fuel storage system, performance characterisation of the torus exhaust processing and design of water detritiation system. High heat flux components have been developed in the framework of ITER R&D programme and based on copper alloy heat sink protected by an armour of beryllium, CFC or tungsten. Studies give an important contribution in defining the nuclear technology programme strategy.

Demo blanket technology R&D results in EU

Abstract The European breeding blanket R&D for DEMO is focussed on the development of two concepts: the Water Cooled Lithium Lead (WCLL) and the Helium Cooled Pebble Bed (HCPB). Test Blanket Modules (TBMs) based on these two concepts will be tested in ITER during the first 10 years of operation. The EU R&D programme is mainly focussed on the development and characterisation of materials: structural (EUROFER), breeding (Pb–17Li, Li4SiO4, Li2TiO3) and neutron multiplier (beryllium), and manufacturing technologies (powder and solid HIP, joining, tritium permeation barrier). In the paper a short description of the two EU breeding blanket concepts, the main R&D results, the key issues still to be clarified and the TBM test programme in ITER are reported.

Characterization of Erosion Dust and Tritiated Products Inside the JET Vessel After the First Tritium Experiment

An experiment was carried out to characterize the erosion products found in the JET vessel after the first tritium experiment. These products were analyzed for carbon, beryllium, inconel metals and tritium. All these elements were present in airborne particles or deposited dust. The tritium was found as tritiated water vapour, and also strongly associated to the suspended or deposited particles. It was more abundant in fine than in coarse particles. The particulate tritium seems to be almost entirely “insoluble” in a water solution.

Design of Narrow Support Elements for Non Planar Coils of Wendelstein 7-X

The stellarator Wendelstein 7-X is presently under construction and assembly in Greifswald, Germany. One of the main structural elements which have to take the electromagnetic forces of the superconducting coil system are the narrow support elements (NSE). They are placed between the non planar coils and have to take very high compressive forces while relative sliding and tilting of the coils must be allowed. The design has been optimised with regard to a proper load distribution among all support elements taking also into account manufacturing and assembly tolerances. The paper describes the design, analysis and tests which have been carried out for the NSEs

Thermal-Hydraulic Experiments for the NET Divertor

An experimental program has been executed on the water subcooled boiling cooling technique in the frame of a collaboration between the NET Team in Germany, ENEA and SIET laboratories in Italy. Experimental data on the heat transfer and the critical heat flux have been obtained for different test section geometries and varying parametrically the hydraulic conditions. The range of hydraulic parameters considered is relevant for the foreseen geometry and conditions of the high heat flux components of fusion reactors such as the divertor. Different turbulence promoters have been tested and their performance directly compared. Trials have been made to set up a system to detect the incipient water boiling and the onset of burn-out using accelerometers. The test procedure and the experimental results are presented and commented on.

DESIGN MANUFACTURING AND TESTING OF THE MONOBLOCK DIVERTOR

Significant progress has been achieved in the design and technological development of the monobloc divertor concept, consisting of metallic tubes inside armour tiles. The materials used were commercial carbon fibre composites for the armour and molybdenum and copper alloys for the tubes. A suitable joining technique between the armour and the tube has been developed and actively cooled mock-ups have been manufactured and tested. Test results show the possibility to withstand over 1000 thermal cycles at a heat flux of 15 MW/m2; the lack of damage to the tiles demonstrates the suitability of the concept in general for high heat flux components.

Doses Due to Accidental Releases of Tritium and Activation Products into the Atmosphere and the Aquatic Environment

Abstract Three different accident consequence assessment codes have been applied to assess the dose to the most exposed individual for unit releases of tritium and activated AISI-316 steel into the atmosphere and the aquatic environment. Co-60 has been identified, within both release scenarios, to be the most hazardous of the investigated radionuclides. Whereas no prevailing exposure pathway could be identified for the airborne releases, the ingestion pathways dominate the dose for the releases into rivers. Doses from 1 gram unit releases of activated steel and tritium in HTO form give higher individual and collective doses (about one order of magnitude) for tritium. Doses from the aquatic releases are lower than those from the atmospheric releases by about two orders of magnitude. But simplifications in the water release scenario indicate that the calculated values have a preliminary character. Probabilistic assessments, taking into account for the differing environmental conditions pertaining at the time of an accident, estimate the variability of the dose values.

Preparation of the Wendelstein 7-X commissioning

Assembly of the superconducting stellarator Wendelstein 7-X is well advanced, and commissioning of the device is being prepared. A first draft of the commissioning tasks has been developed and will be discussed in this paper.