Fusion materials development program in the broader approach activities (original) (raw)
Related papers
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
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.
ITER R & D on breeder blanket materials
Journal of Nuclear Materials, 1991
The International Thermonuclear Experimental Reactor (ITER) activity includes specific R&D in critical areas to support the ITER design. Several of the critical tasks defined as part of the ITER validating R&D are materials related. This paper presents a summary of the breeder/blanket materials related R&D conducted by the four participating parties; viz., European Community, Japan, Soviet Union and the United States. The current effort includes several subtasks in each of the following four task areas: (i) ceramic breeder materials, (2) lead-lithium breeder, (3) aqueous salt solutions, (4) neutron multiplier materials. The objective and scope of each subtask is stated. Although this work was initiated in late 1988, important results have been obtained in all four areas. The work on ceramic breeders includes properties measurements, compatibility studies, and tritium recovery experiments with Liz0 and selected ternary ceramics. Investigations on the lead-lithium breeder have focused primarily on properties, compatibility, and tritium extraction from the 17Li-83Pb eutectic alloy. The work on the aqueous salt breeders includes corrosion, solution chemistry, and radiolysis effects in a radiation environment. Investigations defined under the neutron multiplier task are focused on beryllium; however, lead in the lithium lead alloy also serves as a neutron multiplier. The schedule and planned future work in each area are summarized.
Fusion nuclear technology and materials: status and R&D needs
Fusion Engineering and Design, 1994
The importance of fusion nuclear technology has grown in recent years due to advances in large tokamaks and a strong international movement towards a "next step" device such as ITER. Programs around the world have made major advances in the development of in-vessel reactor components. However, much further R&D will be required in order to develop attractive DEMO components. The realization that testing in a next-step fusion test reactor may be required in as little as 10 years has provided a new sense of urgency. In this paper, the current status and R&D needs are surveyed for the key in-vessel components, including blankets, plasma-facing components, and tritium systems. Special needs in the areas of neutronics, materials development and safety are highlighted. 0920-3796/94/$07.00
Status of the EU DEMO breeding blanket manufacturing R&D activities
Fusion Engineering and Design, 2020
The realization of a DEMOnstration Fusion Power Reactor (DEMO) to follow ITER, with the capability of generating several hundred MW of net electricity and operating with a closed fuel-cycle by 2050, is viewed by Europe as the remaining crucial step towards the exploitation of fusion power. The EUROfusion Consortium, in the frame of the European Horizon 2020 Program, has been assessing four different breeding blanket concepts in view of selecting the reference one for DEMO. This paper describes technologies and manufacturing scenarios developed and envisaged for the four blanket concepts, including nuclear "conventional" assembly processes as GTAW, electron beam and laser welding, Hot Isostatic Pressing (HIP), and also more advanced (from the nuclear standpoint) technologies as additive manufacturing techniques. These developments are performed in conformity with international standards and/or design/manufacturing codes. Topics as the metallurgical weldability of EUROFER steel and the associated risks or the development of appropriate filler wire are discussed. The development of protective W-coating layers on First Wall, with Functionally Graded (FG) interlayer as compliance layer between W and EUROFER substrate, realized by Vacuum Plasma Spraying method, is also propounded. First layer systems showed promising layer adhesion, thermal fatigue and thermal shock properties. He-cooled mock-ups, representative of the First Wall with FG W/EUROFER coating are developed for test campaigns in the HELOKA facility under relevant heat fluxes. First elements of Double Walled Tubes (DWT) manufacturing and tube/plate assembly for the water cooled concept are given, comprising test campaign aiming at assessing their behaviour under corrosion. In addition, further development strategies are suggested.
Solid breeder blanket design and tritium breeding
Fusion Engineering and Design, 1991
Thermonuclear D-T power plants will have to be tritium self-sufficient. In addition to recovering the energy carried by the fusion neutrons (about 80% of the fusion energy), the blanket of the reactor will thus have to breed tritium to replace that burnt in the fusion process. This paper is an attempt to cover in a concise way the questions of tritium breeding, and the influence of this issue on the design of, and the material selection for, power reactor blanket relying on the use of solid breeder materials. Tritium breeding requirements-to breed one tritium per fusion nentron-are shown to be quite demanding. To meet them, the blanket must incorporate, in addition to a tritium breeding lithium compound, a neutron multiplier so as to compensate for neutron losses. Presently prefered lithium compounds are Li20, LiA102, Li2ZrO3, Li4SiO 4. The neutron multiplier considered in most design concepts is beryllium. Furthermore, the blanket must be designed with a view to minimizing these neutron losses (search for compactness and high coverage ratio of the plasma while minimizing the mount of structures and coolant). The design guidelines are justified and the technological problems which limit their implementation are discussed and illustrated with typical designs of solid breeder blanket.