Y. Poitevin

Survey of in-vessel candidate materials for fusion power plants - the european materials R&D programme

The long term European materials RD programme is mainly focused on the development of the following DEMO relevant materials: structural (RAFM steel called EUROFER), plasma facing (W alloy), breeder and neutron multiplier (Pb-17Li, Li4SiO4, Li2TiO3 and Be). Materials like ODS and SiCf/SiC composite are also investigated for advanced fusion reactor concepts. In addition, the materials programme includes activities in nuclear data, modelling of irradiation effects and International fusion materials irradiation facility (IFMIF) (the intense fusion relevant neutron source for material testing). A short description of the EU materials development programme strategy, the ongoing RD activities and the main results obtained so far are reported in this paper

Power plant conceptual study - WCLL concept

Abstract A power plant conceptual study (PPCS) has been launched in the framework of the EU fusion program the objective of which is to demonstrate the credibility of fusion power plant design and the claims for safety and environmental advantages and for economic viability of fusion power. A generic set of requirements, addressing in particular safety, operational and economic aspects, has been set out with inputs from industry and from utilities. Four reactor models have been identified for a complete evaluation. The model which is presented in this paper is based on little extrapolation on both physics and technology, using a water cooled divertor based on ITER technology and associated to the water cooled lithium lead (WCLL) blanket.

In-vessel component designs for a self-cooled lithium-lead fusion reactor

Abstract This paper describes an advanced fusion power reactor based on advanced plasma physics assumptions and large technological extrapolation compared with present-day knowledge. In-vessel components are based on the use of SiCf/SiC composites structure and the use of high-temperature Pb–17Li both as coolant and breeder for the blanket. The use of high-temperature super-conducting coils is also expected. A net electrical output of 1500 MWe is assumed. Tritium breeding self-sufficiency and acceptable operating parameters have been obtained. The resulting conceptual power plant has the potential for a thermal efficiency as high as 61% and for reaching good safety standard. The technological extrapolations assumed in this study provide an indication of the necessary R&D for addressing the most critical issues.

On the use of tin–lithium alloys as breeder material for blankets of fusion power plants

Abstract Tin–lithium alloys have several attractive thermo-physical properties, in particular high thermal conductivity and heat capacity, that make them potentially interesting candidates for use in liquid metal blankets. This paper presents an evaluation of the advantages and drawbacks caused by the substitution of the currently employed alloy lead–lithium (Pb–17Li) by a suitable tin–lithium alloy: (i) for the European water-cooled Pb–17Li (WCLL) blanket concept with reduced activation ferritic–martensitic steel as the structural material; (ii) for the European self-cooled TAURO blanket with SiCf/SiC as the structural material. It was found that in none of these blankets Sn–Li alloys would lead to significant advantages, in particular due to the low tritium breeding capability. Only in forced convection cooled divertors with W-alloy structure, Sn–Li alloys would be slightly more favorable. It is concluded that Sn–Li alloys are only advantageous in free surface cooled reactor internals, as this would make maximum use of the principal advantage of Sn–Li, i.e., the low vapor pressure.

Status of the design and performances of the WCLL test blanket module for ITER-FEAT

Abstract The water-cooled lithium–lead (WCLL) DEMO blanket is one of the two European blanket lines to be further developed with the aim, in the short term, of manufacturing a DEMO blanket mock-up, named the test blanket module (TBM) to be tested in ITER-FEAT. This blanket line is based on reduced-activation ferritic-martensitic steel of the 9% Cr class as structural material, the liquid alloy Pb–17Li as breeder and neutron multiplier and water at typical PWR conditions as coolant. The WCLL TBM is expected to use all technologies required for that concept and is essentially formed by a directly cooled steel box having the function of Pb–17Li container and by a double-walled C-shaped tube bundle, immersed in the liquid metal, in which the water coolant circulates. The design of the TBM and its ancillary circuits is described. Conceptual studies have shown that, in the range of the foreseen heat flux and neutron wall loading in ITER-FEAT, DEMO relevant testing conditions could be reached with a unique reference TBM design by properly adjusting the thermal–hydraulic conditions in both the breeder zone and segment box. A tentative test plan in ITER has been established with the overall objective to evaluate the TBM behavior (e.g. functional, thermal, mechanical, neutronic) in conditions which are, as much as possible, relevant to DEMO.

Further improvements of the water-cooled Pb-17Li blanket

Abstract The water-cooled lithium–lead (WCLL) blanket is based on reduced-activation ferritic–martensitic steel as the structural material, the liquid alloy Pb–17Li as breeder and neutron multiplier, and water at typical PWR conditions as coolant. It was developed for DEMO specifications and shall be tested in ITER. In 1999, a reactor parameter optimization was performed in the EU which yielded improved specifications of what could be an attractive fusion power plant. Compared to DEMO, such a power reactor would be different in lay-out, size and performance, thus requiring to better exploit the potential of the WCLL blanket concept in conjunction with a water-cooled divertor. Several new approaches are currently under evaluation. This paper outlines several specific modifications, it highlights progress made on various issues and outlines the R&D work which is still required to define an improved reference design for the WCLL concept.

Neutronic analysis for the EU water-cooled lithium–lead test blanket module in ITER-FEAT

Comprehensive neutronic analyses using the MCNP Monte Carlo code have been performed to predict the neutronics performance of the EU water-cooled lithium–lead test blanket module (TBM) integrated in ITER-FEAT. The analyses have been performed using models representing the complex ITER and TBM structure close to reality. Tritium generation and parameters relevant to the lifetime performance of the TBM such as helium, hydrogen and atomic displacement production have been calculated. Assuming 90% 6Li enrichment and 22% duty cycle, the estimated tritium production amounts to 3.8738 × 1016 T atoms s−1 (16.62 mg d−1). The maximum radiation damage obtained, based on 0.3 MWa m−2 neutron fluence, is in the first wall first steel layer of the breeder container totalling 2.77 dpa of Fe in Euro steel, 19.3 appm He production and 103 appm H generation. The calculated helium production in the demountable hydraulic connections of TBM is below the required limit for reweldability of steel structure.

Potential and limits of water-cooled Pb–17Li blankets and divertors for a fusion power plant

Abstract Blankets and divertors are key components of a fusion power plant. They have a large impact on the overall plant design, its performance and availability, and on the cost of electricity. The water-cooled Pb–17Li (WCLL) blanket uses reduced activation ferritic–martensitic steel as structural material. It was previously validated under numerous aspects such as TBR, mechanical and thermo-mechanical stability, thermal–hydraulics, MHD, safety and others. This was done assuming the specifications for a European DEMOnstration reactor which were fixed back in 1989. A WCLL blanket would best be combined with a water-cooled divertor so that a single coolant could be used for the entire reactor. Several divertor designs were proposed recently. This paper investigates the applicability of the WCLL blanket concept and a water-cooled divertor in attractive power reactors with increased power densities compared with DEMO.