ITER thermal shields at the starting phase of procurement (original) (raw)

Progress of detailed design and supporting analysis of ITER thermal shield

2010

The detailed design of ITER thermal shield (TS), which is planned to be procured completely by Korea, has been implemented since 2007. In this paper, the design and the supporting analysis are described for the critical components of the TS, the vacuum vessel TS (VVTS) outboard panel, labyrinths and VVTS supports. The wall type of VVTS outboard panel was changed from double wall to single wall, and the verification analyses were carried out for this design change. The dimensions of the labyrinths were determined and the heat load through the labyrinth was analyzed to check the design requirement. The preliminary result of the VVTS inboard and outboard supports were obtained considering the structural rigidity.

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.

European development of prototypes for ITER high heat flux components

Fusion Engineering and Design, 2000

The extensive EU research and development, on international thermonuclear experimental reactor (ITER) high heat flux (HHF) components aims at the demonstration of prototypes for the divertor and baffle with challenging operating requirements. The recent progress of this development is summarised in the paper, particularly concerning the manufacture and testing of mock-ups and prototypes. The available results demonstrate the feasibility of robust solutions with carbon and tungsten armour.

Design approach of the vacuum vessel and thermal shields towards assembly at the ITER-site

Fusion Engineering and Design, 2009

Recent progress of the ITER vacuum vessel (VV) and thermal shields (TS) design is presented. As the ITER construction phase approaches, the design of the VV and TS (in particular, the vacuum vessel TS-VVTS) has been improved and developed in more detail with the focus on better performance, improved manufacturing ability and successful assembly at the ITER-site. In addition to the design progress, the main principles and operations for assembly of the VV, VVTS and other TS components at the ITER-site are described.

EU contribution to the procurement of blanket first wall and divertor components for ITER

Fusion Engineering and Design, 2018

Fusion for Energy (F4E), the European Domestic Agency for ITER, is responsible for a significant share of the overall procurement of internal components inside the vacuum vessel, the largest ones being as follows: (1) 48% of the ITER blanket first wall (FW), (2) the divertor cassette bodies and (3) the divertor inner vertical target. Procurement strategies have been implemented by the In-Vessel Project Team at F4E aiming at mitigating technical and commercial risks for the procurement of these ITER internal components, promoting as far as possible competition among industrial partners. These procurement strategies have been supported by extensive research and development (R&D) work programmes, implemented over more than 15 years in Europe, to develop various fabrication technologies especially for plasma facing components. They include in particular the manufacture and testing of small-scale mock-ups, medium-scale mock-ups and full-scale prototypes of blanket FW panels and divertor inner vertical targets. In these R&D work programmes, significant efforts have been devoted to the development of reliable armour-to-heat sink materials joining techniques. More recently, fullscale prototypes of the divertor cassette body have also been fabricated. This paper presents the procurement strategies implemented by F4E for the supply of the European contribution to the procurement of the above ITER internal components, together with the main outcome of the R&D work programmes and the state of progress of the respective qualification programmes.

Thermal–mechanical test on ITER primary first wall mock-ups

Fusion Engineering and Design, 2002

In 1998, in the frame of the ITER EDA phase, an European R&D Programme for the Blanket Design was implemented for developing and selecting the materials and the relevant fabrication procedures for manufacturing the shielding modules of the ITER Primary Wall. The fabrication of several Beryllium armored small scale mock-ups, reproducing representative portions of a Primary Wall panels, was also launched (Fusion Technol. (1998) 195). Further experimental activities were also programmed for investigating the thermal Á/mechanical behavior of these mock-ups at high heat flux and under thermal fatigue tests. In 2001, the ITER European Home Team decided to assign to ENEA a contract for the thermal fatigue testing of six mock-ups aiming at verifying the reliability of the Beryllium/Dispersion Strengthened Copper alloy/Stainless Steel and Beryllium/Precipitation hardened Copper alloy/Stainless Steel joints manufactured by solid Hot Isostatic Pressing (HIP) procedure (Technical Specification for the Thermal Fatigue Tests of Be protected EDA Mock-ups). The paper presents the results of the FEM thermal Á/mechanical analyses performed by ANSYS code and the progress of the first test campaign.

European achievements for ITER high heat flux components

Fusion Engineering and Design, 2001

This paper summarises the main activities carried out by the EU Home Team to develop suitable solutions for the ITER high heat flux components, namely the divertor, the baffle and the limiter. The available results demonstrate that the EU have the capability to manufacture high heat flux components with carbon fibre reinforced carbon, tungsten and beryllium armours which all exceed the ITER design requirements.

Computational thermo-fluid exploratory design analysis for complex ITER first wall/shield components

Fusion Engineering and Design, 2008

Engineers in the ITER US Party Team used several computational fluid dynamics codes to evaluate design concepts for the ITER first wall panels and the neutron shield modules. The CFdesign code enabled them to perform design studies of modules 7 and 13 very efficiently. CFdesign provides a direct interface to the CAD program, CATIA v5. The geometry input and meshing are greatly simplified. CFdesign is a finite element code, rather than a finite volume code. Flow experiments and finite volume calculations from SC-Tetra, Fluent and CFD2000 verified the CFdesign results. Several new enhancements allow CFdesign to export temperatures, pressures and convective heat transfer coefficients to other finite element models for further analysis. For example, these loads and boundary conditions directly feed into codes such as ABAQUS to perform stress analysis. In this article, we review the use of 2-and 4-mm flow driver gaps in the shield modules and the use of 1-mm gaps along the tee-vane in the front water header to obtain a good flow distribution in both the first wall and shield modules for 7 and 13. Plasma heat flux as well as neutron heating derived from MCNP calculations is included in the first wall and shield module analyses. We reveal the non-uniformity of the convective heat transfer coefficient inside complex 3D geometries exposed to a one-sided heat flux and non-uniform volumetric heating. Most models consisted of 3-4 million tetrahedron elements. We obtained temperature and velocity distributions, as well as pressure drop information, for models of nearly exact geometry compared to the CATIA fabrication models. We also describe the coupling to thermal stress analysis in ABAQUS. The results presented provide confidence that the preliminary design of these plasma facing components will meet ITER requirements.

Russian development of enhanced heat flux technologies for ITER first wall

Fusion Engineering and Design, 2012

Recently the ITER first wall (FW) design has been significantly upgraded to improve resistance to electromagnetic loads, to facilitate FW panel replacement and to increase FW ability to withstand higher (up to 5 MW/m 2) surface heat loads. The latter has made it necessary to re-employ technologies previously developed for the now-abandoned port limiters. These solutions are related to the cooling channel with CuCrZr-SS bimetallic walls and hypervapotron type cooling regime, optimization of Be-tiles dimensions and Be to CuCrZr joining technique. A number of representative mockups were tested at high heat flux (HHF) at the Tsefey electron-beam facility to verify the thermo-hydraulic characteristics of the reference cooling channel design at moderate water flow velocities (V = 1-3 m/s, P = 2-3 MPa, T = 110-170 • C). The heat flux was gradually varied in the range of 1-10 MW/m 2 until the critical heat flux was registered. The mockups of hypervapotron structure demonstrated the required cooling efficiency and critical heat flux margin (1.4) at a water velocity of ≥2 m/s. Dimensions of Be armor tiles strongly affect the thermo-mechanical stresses both in the CuCrZr cooling wall and at the Be-CuCrZr interface. Results of tile dimensions optimization (variable in the range 12 mm × 12 mm × 6 to 50 mm × 50 mm × 8 mm) obtained by the HHF (variable in the range of 3-8 MW/m 2) experiments are presented and compared with analysis. It is shown that optimization of the tile geometry and joining technology provides the required cyclic fatigue lifetime of the reference FW design.