Progress and Achievments of the ITER L-4 Blanket Project (original) (raw)
Related papers
Engineering design of the ITER blanket and relevant research and development results
Fusion Engineering and Design, 1999
The design of the ITER blanket is presented together with the related technology which has been developed. The evolution of this component since the beginning of the EDA is explained in relation to the developing understanding of the thermal deformations and of the electromagnetic forces. These loads lead to a system composed of compact modules protecting a continuous support shell called a backplate. The backplate is a stiff double wall construction which conveys the coolant to the modules. The supports of the module are flexible and allow relative thermal expansions. They are connected and disconnected to the backplate by bolts operated through holes in the front face of the module. The coolant connections and the electrical straps located on the back of the modules are reached similarly. The first wall is integral with the module and cooled in series. A research and development program on materials and joining methods defined the construction path which has been tested in prototypes. The main body is built of stainless steel by forging and drilling or powder hot isostatic pressing (HIP), depending on the complexity of the shape. The first wall includes a dispersion strengthened copper heat sink which is hot isostatic pressed onto the steel body. Beryllium is the basic plasma facing material and is attached by HIP to the copper. Prototypes of the module attachment have been built and are under integrated tests.
Status of the EU R&D programme on the blanket-shield modules for ITER
Fusion Engineering and Design, 2008
A research and development (R&D) programme for the ITER blanket-shield modules has been implemented in Europe to provide input for the design and the manufacture of the full-scale production components. It involves in particular the fabrication and testing of mock-ups and full-scale prototypes of shield blocks and first wall (FW) panels. This paper summarises the main achievements obtained so far and presents the latest results of this R&D programme. In particular, it reports the status of the shield fabrication development programme with the manufacture of a full-scale shield prototype. It also reports the latest results of high heat flux and thermal fatigue tests of FW mock-ups. It describes the preparation of irradiation experiments of Be coated FW mock-ups. Finally, it presents the outline of a possible qualification programme that each contributing participant teams should pass prior to the procurement of the blanket-shield modules for ITER.
ITER-Test blanket module functional materials
Journal of Nuclear Materials, 2007
Solid breeder and liquid breeder blanket concepts are being developed to be tested in ITER. In addition to the fundamental investigations on the use of reduced activation ferritic/ferriticmartensitic steels (RAFMS) or V-alloys as the structural material, tritium breeder and neutron multiplier materials, and corresponding corrosion issues, there are three classes of functional materials being developed. Performance of these functional materials will impact significantly the performance of respective blanket concepts. Two classes of materials are related to the selfcooled liquid breeder design concepts. The first class is MHD coatings like Er 2 O 3 and Y 2 O 3 , which are proposed as the MHD insulation options required for the self-cooled Li breeder concepts. The second class is flow channel inserts (FCI) like SiC f /SiC composite material being developed to perform the functions of MHD and thermal insulation for the dual-coolant PbLi concepts. The third class of functional material is the tritium permeation barrier material. This class of material is commonly needed for solid and liquid breeder blanket concepts to reduce the permeation loss of tritium to the environment. Description and development status of these functional materials are described in this paper.
Progress in the ITER blanket design
17th IEEE/NPSS Symposium Fusion Engineering (Cat. No.97CH36131), 1998
Significant improvements of the ITER blanket were possible with the better definition of the thermal and electromagnetic effects. A supporting double wall backplate with reinforcement about the ports has been defined. The module has been simplified and the intrusion of the connections reduced. The attachment has been assessed by system analyses. The cooling system is configured to simplify the leak detection. 0-7803-4226-7/98/$10.00 0 1998 IEEE
Fusion Engineering and Design, 2020
Two breeding blanket concepts, the Helium-Cooled Lithium-Lead (HCLL) and the Helium-Cooled Pebble-Bed (HCPB), developed in Europe in the frame of DEMO and relevant TBM studies, present many similarities in terms of design and manufacturing. As an example, all structure sub-components are internally cooled by helium circulating in meandering squared section channels. Several technologies have been investigated in the frame of EU fusion research programme for the manufacturing, including a specialized machining, fusion and diffusion welding and joints inspection, of DEMO breeding blanket and TBM sub-components (Cooling Plates (CP), Side Caps (SCs), Stiffening Plates (SPs), First Wall (FW)) and their assembly, main of which are discussed in this paper. The manufacturability of DEMO blanket modules Back Supporting Structure (BSS), a big structure located behind blanket modules and supporting them, is also discussed. The applicability of technologies takes into account specificities of EUROFER97 steel, foreseen as structural material. For each assessed technology, main results, e.g. in terms of mechanical properties and microstructure of weld joints, are presented and main advantages and drawbacks are summed up, in order to identify most promising technology/ies and to propose manufacturing scenarios. It should be noted that these developments are performed according to standards and professional codes (RCC-MRx). Further development strategies are also briefly discussed in the paper.
Fusion Engineering and Design, 2000
In the frame of the international thermonuclear experimental reactors (ITER) collaboration, the European home team (EU HT) has committed significant efforts on the R&D for the Shield Blanket. This paper summarises the main achievements of this programme, which have been obtained over the last 7 years. The depth of R&D extends from generic activities up to the manufacture of prototypes, but has, in accordance with the design progress, reached different stages of maturity for the various components. New ITER options being considered since early 1998 have not made these activities irrelevant. With few exceptions, the results are still applicable for less ambitious next step machines, or transferable to components with similar functions or requirements.
Test blanket modules in ITER: An overview on proposed designs and required DEMO-relevant materials
Journal of Nuclear Materials, 2007
Within the framework of the ITER Test Blanket Working Group, the ITER Parties have made several proposals for test blanket modules to be tested in ITER from the first day of H-H operation. This paper gives an overview of the proposed TBMs designs, of the ITER boundary conditions and of the expected TBM operating conditions. Operating conditions will vary throughout the various ITER phases, starting from the initial H-H phase where no neutrons and, therefore, no nuclear volume heating will be present, to the later D-T phase where pulses of up to 3000 s length may be expected. The paper is focused on the design requirements for the materials and subcomponents that will be used in the various TBMs, from the viewpoint of both the materials performance and the required R&D.
Status of ITER blanket attachment design and related R&D
Fusion Engineering and Design, 2013
h i g h l i g h t s • ITER blanket attachment system went through a significant design upgrade and become basically compliant with specified design loads and required cyclic lifetime. • Upgrade of flexible supports allowed the doubling of cross sections of central bolts. Ceramic coatings were relocated to much larger areas on conical pairs screwed into shield blocks. • Key pads were relocated from keys of vacuum vessel into keyways of shield blocks and reshaped to enlarge areas of lateral interfaces with ceramic electro-insulating coatings. • Ceramic coatings are hidden between pads and enclosures in keyways with a purpose to minimize their wear rate, which depends on peak friction stress and cyclic sliding path. • Ceramic coatings to be verified by experiment, with several R&D aimed to collect statistically sufficient data on their reliability and durability in ITER relevant cyclic loading conditions.
Fusion Technology, 1992
This paper presents the recent development studies on the adaptation of the European Ceramic BIT Blanket to updated DEMO specifications. The adaptation work is in progress, since 1990, when a detailed comparison between two existing designs lead to the selection of an unique concept. The main new developments concern the separation in two parts of the inboard blanket segments at the level of the lower divertor, the consequent improvement of the blanket coverage, the simplification of maintenance operations, and finally the increased compactness of the blanket because of the inclusion of the shielding into the breeder assembly.
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