Blankets – key element of a fusion reactor – functions, design and present state of development (original) (raw)
Related papers
Review of blanket designs for advanced fusion reactors
Fusion Engineering and Design, 2008
The dominating fraction of the power generated by fusion in the reactor is captured by neutron moderation in the blanket surrounding the plasma. From this, the efficiency of the fusion plant is predominated by the technologies applied to make electricity or hydrogen from the neutrons. The main blanket concepts addressed in this paper are advanced ceramic breeder concepts, dual coolant blankets as well as self-cooled liquid metal and Flibe blankets. Two important questions that are addressed are: (i) Can we draw a bottom line conclusion on the most promising concept(s)? (ii) What are the common issues to be resolved independently from individual design and layout proposals to define a feasible route towards advanced fusion reactors? For ceramic breeder concepts, a key issue in the long term could be the limitation of beryllium as the considered multiplier in terms of world sources and achievable temperature levels. For liquid metal blankets, attractive long-term visions have been developed but major technological challenges also exist for the in-vessel blanket technology and the corresponding subsystems. The paper proposes a strategic conclusion derived from the review of blanket designs for advanced fusion reactors.
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.
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.
Journal of Nuclear Materials, 1994
The structural materials considered for solid and liquid metal breeder blankets are the austenitic and martensitic steels and vanadium alloys. The principal concerns with these materials are: (a) the high-temperature-induced swelling of the austenitic steels, (b) the low temperature irradiation embrittlement of martensitic steels, and (c) the exact specification of the preferred alloy composition(s), properties during and following irradiation, and technological aspects (fabrication and welding) for the vanadium alloys. Solid breeder blankets are based on the use of lithiated ceramics such as Li,O, LiA102, Li,SiO, and Li,ZrO, and beryllium as a neutron multiplier. The main uncertainty with these materials is their behaviour under irradiation, particularly at higher bumups and fluences than have been achieved hitherto. Liquid metal blankets, utilising pure Li or the LiPb eutectic as the tritium breeding material, can be either self-or separately-cooled; separate coolants include water (with LiPb) and helium. The important materials issues with the LiPb are the development of permeation barriers to contain the tritium and, for the self-cooled option, electrical insulators to reduce the MHD pressure drop to acceptable levels. 0022-3115/94/$07.00 0 1994 Elsevier Science B.V. All rights reserved SSDI 0022-3115(94)00066-W
Reducing the Peak-to-Average-Power-Ratio in Fusion Blankets
Fusion Science and Technology, 2017
Breeding blankets with integrated first wall are one of the most critical components of nuclear fusion reactors. Blankets breeding zones are characterized by steep nuclear heating gradients due to the exothermic nuclear reaction 6 Li(n, α)T and the high intensity neutron flux in the proximity of the first wall. Non-uniformity in nuclear heating can generate sharp temperature gradients that deeply affect material properties. This conceptual study explores an original way to flatten nuclear heating profiles by proposing a blanket characterized by layers of different 6 Li enrichment in the breeder region while maximizing Tritium Breeding Ratio (TBR) and power generation. Two types of fusion blanket are studied: (1) Helium Cooled Ceramic Reflector (HCCR) and (2) Dual Coolant Lead Lithium (DCLL). For HCCR, it is found in the optimal design case, that the power peak-to-average can be reduced by 47.85%, 42.45% and 54.13% in the front, middle and back channel respectively when compared to the reference design. On the other side, we found that this method of profile flattening is not appealing for DCLL, under the geometrical configuration and material selection in this particular blanket design, since most of nuclear heating is caused by photon heat deposition.
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.
Overview of Fusion Blanket R&D in the Us Over the Last Decade
2005
We review here research and development progress achieved in US Plasma Chamber technology roughly over the last decade. In particular, we focus on two major programs carried out in the US: the APEX project (1998-2003) and the US ITER TBM activities (2003-present). The APEX project grew out of the US fusion program emphasis in the late 1990s on more fundamental science and innovation. APEX was commissioned to investigate novel technology concepts for achieving high power density and high temperature reactor coolants. In particular, the idea of liquid walls and the related research is described here, with some detailed examples of liquid metal and molten salt magnetohydrodynamic and free surface effects on flow control and heat transfer. The ongoing US ITER Test Blanket Module (TBM) program is also described, where the current first wall / blanket concepts being considered are the dual coolant lead lithium concept and the solid breeder helium cooled concepts, both using ferritic steel...
Blanket design using FLiBe in helical-type fusion reactor FFHR
Journal of Nuclear Materials, 1997
The blanket design for a force-free helical reactor (FFHR) is presented, which is a demo-relevant heliotron-type D-T fusion reactor based on the first all-superconducting-coils device, LHD (large helical device) under construction in NIFS at present. For the goal of a self-ignited reactor of 3 GW thermal output, the design parameters at the first stage for concept definition of FFHR have been investigated. The main feature of FFHR is a force-free-like configuration of helical coils, which makes it possible to simplify the coil supporting structure and to use a high magnetic field instead of high plasma beta. The other feature is the selection of molten-salt FLiBe as a self-cooling tritium breeder for mainly safety reasons owing to the low tritium inventory, low reactivity with air and water, low pressure operation, and low MHD resistance compatible with a high magnetic field. In particular, as common issues in fusion reactors, the FLiBe blanket system in FFHR is expressed in detail by showing engineering possibilities to overcome key issues on tritium permeation, material corrosion, heat transfer, operation pressure, etc. The basic design for maintenance and repair of the blanket is also discussed. © 1997 Elsevier Science B.V.
Fusion Engineering and Design, 2000
Among the concepts considered in the Advanced Power Extraction (APEX) study is the Hecooled refractory metal FW and blanket concept. Refractory metals exhibit high operating temperature and can offer good capability for withstanding high power density operation that is the focus of the APEX study. In this paper, we assess the impact of using various refractory metal on the nuclear heating profiles across the blanket and power multiplication, PM, and on the tritium breeding profiles and tritium breeding ratio, TBR. The refractory metals considered with liquid lithium breeder are W, TZM, and Nb-1Zr. The impact of Li-6 enrichment on these profiles and on TBR and PM is also assessed. Comparison of these nuclear characteristics is also made to other liquid breeder (Flibe and Li-Sn). Because the moderation power of these breeders to neutron energy varies among them, the damage to the structure is different with various structure/breeder combinations. The damage parameters (DPA, helium and hydrogen production)
HIP technologies for fusion reactor blankets fabrication
Fusion Engineering and Design, 2000
The benefit of HIP techniques applied to the fabrication of fusion internal components for higher performances, reliability and cost savings are emphasized. To demonstrate the potential of the techniques, design of new blankets concepts and mock-ups fabrication are currently performed by CEA. A coiled tube concept that allows cooling arrangment flexibility, strong reduction of the machining and number of welds is proposed for ITER IAM. Medium size mock-ups according to the WCLL breeding blanket concept have been manufactured. The fabrication of a large size mock-up is under progress. These activities are supported by numerical calculations to predict the deformations of the parts during HIP'ing. Finally, several HIP techniques issues have been identified and are discussed.