Claudio Nardi

Development of the EU water-cooled Pb-17Li blanket

The reference concept of the EU water-cooled Pb-17Li DEMO blanket is essentially formed by a directly-cooled steel box having the function of Pb-17Li container and by a double-wall U-tube bundle, immersed in the liquid metal, in which the pressurised water-coolant flows. The structural material is martensitic steel. All blanket performances satisfy DEMO requirements, such as tritium breeding self-sufficiency, capability of the box-structures to withstand water-coolant pressure under faulted conditions, limitation of the Pb-17Li/steel interface temperature to 480°C, power conversion efficiency of about 35%, and acceptable increase of temperature in case of out-of-vessel LOCA. A preliminary design of the test module and of the associated circuit components to be tested in ITER has been performed taking into account ITER specifications for test ports and ITER pulsed working conditions. Test-module DEMO-relevance has been one of the leading criteria for the module design. Preliminary module manufacturing sequences have been defined. The paper recalls the associated R and D activities which have to be finalised prior to the ITER testing and mainly involve tritium permeation barrier development, tritium extraction from Pb-17Li, advanced manufacturing techniques, definition of water/Pb-17Li interaction counter-measures, and Li-concentration control techniques.

Comparative availability analysis of the four European DEMO blanket concepts in view of the selection exercise

Abstract The four blanket concepts developed within the European Fusion Technology program and supported by the European Community are assessed from the reliability/availability point of view. The reliability of blanket segments influences the availability of the whole plant. In case of a failure, the affected blanket segments must be repaired or exchanged. The mean down time necessary for segments repair or exchange is an important factor for the overall plant availability. It has been assumed to be 3 months for all concepts. To allow a meaningful comparative assessment, an identical evaluation scheme was applied to all concepts using the same basic data. The results show that differences in the availability of blanket concepts are relatively small, with availabilities ranging between 84.3 and 87.7%. The absolute numbers, however, are sensitive to assumptions and to failure modes taken into consideration.

High-temperature residual strain measurements, using neutron diffraction, in brazed Cu/CFC graphite divertor structures

Abstract This work presents a study of residual strain evolution in a three layer castelled mock-up for ITER obtained brazing carbon fibre composite (CFC) graphite with dispersion strengthened Cu (and with a GLIDCOP interlayer). Neutron diffraction measurements were carried out, on graphite at 30°C, 300°C and 600°C and on dispersion strengthened (DS) Cu at 600°C. For each material, the reference, unstrained value of the lattice parameter was obtained by measuring a powder of the same material at the same temperatures. While in DS Cu negligible strain values are found at high temperature, a more complicated evolution of the strains is observed in CFC graphite relating to sample geometry.

A breeding blanket in ITER-FEAT

Abstract At the beginning of life ITER-FEAT will be equipped with a shielding blanket; nevertheless since many advantages can be envisaged in having a breeding blanket in the machine, such a concept has been investigated as well. It is a water-cooled blanket in which both ceramic breeder and beryllium multiplier are present in the form of single sized pebble beds. The lithium ceramic breeder is in contained in cylindrical tubes surrounded by the beryllium pebble bed. For several years, studies have been devoted to observing the heat transfer in pebble beds. Since the experimental activities aiming at estimating the relationships between thermal parameters, temperature and stresses in the pebble beds in relevant temperature range have not been completed yet, a sensitivity analysis has been performed in order to verify the variations of breeder and multiplier temperatures in the breeding blanket by assuming different thermal conductivity for the beryllium pebble bed. The thermal analyses performed show that a breeding blanket in ITER-FEAT can be operated by keeping minimum temperature of the breeder and maximum temperature of the multiplier within allowable limits. Neutronic calculations indicate the possibility of having a breeder blanket with a significant TBR in ITER-FEAT.

High temperature residual strain measurements in a brazed sample for NET/ITER

The strain-free temperature in a brazed divertor mock-up for NET/ITER was measured using neutron diffraction. The investigated sample was a CuCrZr alloy cooling pipe armoured by a carbon/carbon fiber composite (CFC) monoblock tile. The joining between CFC and CuCrZr alloy was obtained by an intermediate layer of copper. Neutron diffraction measurements were carried out measuring the strain evolution as a function of temperature in the CuCrZr pipe and the lattice parameter of a reference powder of the same material, in the same temperature range. Both radial and tangential strains were determined inside the pipe in two representative points at 90°, finding well-reproducible results namely: radial strain vanishes at 370°C and tangential strain at 430°C approximately. These experimental results are discussed in the light of numerical analyses and simplified finite element method (FEM) calculations.

Objectives feasibility assessment of the water-cooled lithium–lead mock-up testing in ITER

Abstract This paper presents a short description of the latest design version of the water-cooled Pb–17Li test-blanket module system and the proposal for the test program to be performed during the ITER Basic Performance Phase. The present R&D program foresees to start the test on the first day of ITER operation. The test-blanket module is expected to use all technologies required for the corresponding DEMO blanket (e.g. same structural material, double-wall tubes, permeation barriers) presently under development. Main testing objectives are the demonstration of the system functionality and the validation of the predictions obtained from the theoretical analyses. The feasibility of such objectives is discussed from the point of view of instrumentation and interpretation.

Design development and manufacturing sequence of the European water-cooled Pb-17Li test blanket module

Abstract In 1996, the European Community started the development of a water-cooled Pb-17Li blanket test module for ITER. First tests are currently scheduled to start with the beginning of the Basic Performance Phase prior to d–t operation. The test module is designed to be representative for a DEMO breeding blanket and relies on the liquid alloy Pb-17Li as both tritium breeder and neutron multiplier material, and water at PWR pressure and temperature as coolant. The structural material is martensitic steel. The straight, box-like structure of this blanket confines a pool of liquid Pb-17Li which is slowly circulated for ex-situ tritium extraction and lithium adjustment. The box and the Pb-17Li pool are separately cooled, the former with toroido-radial tubes, the latter with a bundle of double-walled U-tubes, equally made of martensitic steel and equipped with a permeation barrier. This paper presents the latest design and three manufacturing schemes with different degrees of technology. Advanced techniques such as solid or powder HIP are proposed to provide design flexibility. With a 3D neutronics analysis, the power and tritium generation were determined.