Fusion reactor blanket with Li17Pb83 eutectic (original) (raw)

Conceptual Study of a Lithium Lead Eutectic Blanket for a Power Reactor

Nuclear Technology - Fusion, 1983

In the frame of the recent CEA studies aiming at the evaluation and the comparison of various candidate blanket concepts in view of their possible extrapolation to the commercial power reactor, the present work examines the potential interest of a 15 MPa pressurized water cooled Li]jPb83 blanket. After a brief presentation of the main reactor parameters, the body of the paper is devoted to the engineering optimization of the blanket arrangement, in terms of tritium breeding (minimization of the water content), coolant manifolding (minimum coolant cross section, minimum number of connections and easy access for maintenance) and adaptation to the steep power and irradiation gradients, typical of Li 17 Pb 8 o and crucial for a power reactor. Polo'idal cooling direction, long heated length and segmentation into the radial direction (breeder rows) provide some answers to these preoccupations and could be recommended for the next step liquid blanket studies, in order to anticipate the requirements of the commercial reactor.

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

Fusion Engineering and Design, 2000

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.

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

Fusion Engineering and Design, 2001

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.

Candidate blanket concepts for a European fusion power plant study

Fusion Engineering and Design, 2000

The breeding blanket is an essential in-vessel component for fusion power plants based on the deuterium -tritium fuel. It is submitted to severe operating conditions such as high surface heat flux on the first wall ( \ 0.5 MW/m 2 ) and to very high 14 MeV neutron flux (\10 18 n/m 2 s). Because of the simultaneous requirement of very demanding performances, such as tritium breeding self-sufficiency, high thermal cycle efficiency, high availability and high safety standards, the number of candidate breeding blanket concepts is limited. After recalling advantages and drawbacks of possible combinations of structural materials, coolants, and breeder materials, this paper summarizes the characteristics and the performances of some potential candidate concepts for a European power plant study and associated required R&D.

European Blanket Development for a DEMO Reactor

Fusion Technology, 1994

There are four breeding blanket concepts for a fusion DEMO reactor under development within the framework of the fusion technology programme of the European Union (EU). This paper describes the design of these concepts, the accompanying R + D programme and the status of the development.

Status of Fusion Reactor Blanket Evaluation Studies in France

Fusion Technology, 1985

In the frame of recent CEA studies aiming at the evaluat ion and at the comparison of various candidate blanket concepts in moderate power conditions (P n^2 MW/m 2), the present work examines the neutronic and thermomechanical performances of a water cooled Lij7Pb83 tubular blanket and those of a helium cooled canister blanket taking advantage of the excellent breeding capability of composite Beryllium / LiA10 2 (85/15%) breeder elements. The purpose of the following discussion is to justify the impetus for these reference concepts and to summarize the state of their evaluation studies updated by the continuous assimilation of calculations and experiments in progress.

Enhancing tritium breeding characteristics of a fusion reactor blanket

Annals of Nuclear Energy, 1998

In this paper we have reported our results of a detailed time dependent study of neutron spectra, tritium production rate, tritium breeding ratio, slowing down time and average energy following a pulse of 14MeV neutrons in Li + C assemblies of different sizes. We have solved the time dependent diffusion equation with energy dependent buckling as an eigenvalue problem. Our results show that an initial 14 MeV pulse slows down to the keV region within about 200 ns after its injection. At later times, its energy decreases rather slowly because elastic scattering is the sole mechanism of energy loss. Since most of the tritium producing reactions take place with neutrons having energies in this region, the tritium production rate is higher in a Li + C system than a corresponding natural lithium assembly. Moreover, in Li + C the calculated value of TBR is almost twice the value obtained for a natural lithium assembly of the same size. Also, by varying the concentration of 6Li we find that the value of TBR tends to saturate beyond about 40%.