Engineering design of the ITER blanket and relevant research and development results (original) (raw)

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

ITER blanket manifold final design and validation

Fusion Engineering and Design, 2017

h i g h l i g h t s • The Blanket Manifolds provide cooling water to the plasma facing Blanket. • The design, verification by analysis and practical validation have been finalised. • Good thermal conductance and electrically insulated results in conflicting needs. • Requirements to customise to suit as built Torus and remote handling maintenance.

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.

Progress and Achievments of the ITER L-4 Blanket Project

1999

The L-4 Blanket Project embraces the R&D of the ITER Shielding Blanket, and its main objective is the fabrication of prototype components. This paper summarises the main conclusions from the materials R&D and the development of technologies which were required for the prototype specifications and manufacturing. The main results of the ongoing testing activities, and of the component manufacture are outlined. The main objectives of the project have been achieved including improvements of the material properties and of joining technologies, which resulted in good component quality and high performance in qualification 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 first wall/shield blanket

Fusion Engineering and Design, 1998

The blanket system comprises first wall/shield modules which are supported on a structural shell called a Back Plate containing water channels for cooling the modules. A modular design of the first wall/shield was chosen to allow independent maintenance through the horizontal mid-plane ports using in-vessel remote handling equipment. The modules are mechanically attached to the back plate for assembly and maintenance. The first wall/shield modules can be replaced with breeding modules for the EPP. In the first wall (FW) region, the water is contained in 316 LN stainless steel pipes surrounded by a copper heat sink except in high heat flux regions where copper pipes with a stainless steel liner are used in the limiter and baffles. The plasma facing surface of the FW will be beryllium except for the lower region of the baffles, where tungsten is used. Electromagnetic analysis and structural analysis has been performed on the back plate and FW/shield modules during a plasma disruption and for a VDE with toroidally asymmetric halo currents. A major R and D project is being conducted to provide needed input for design of the blanket system and to confirm the fabrication technology.

Thermal–hydraulic and thermo-structural analysis of first wall for Indian DEMO blanket module

Fusion Engineering and Design, 2009

The first wall (FW) is one of the most important components of any fusion blanket design. India has developed two concepts of breeding blanket for the DEMO reactor: the first one is Lead-Lithium cooled Ceramic Breeder (LLCB), and the second one is Helium-Cooled Ceramic Breeder (HCCB) concept. Both the concept has the same kind of FW structure. Reduced Activation Ferritic Martensitic steel (RAFMS) used as the structural material and helium (He) gas is used to actively cool the FW structure. Beryllium (Be) layer of 2 mm is coated on the plasma side of the FW as the plasma facing material. Cooling channels running in radial-toroidal-radial direction in the RAFMS structure are designed to withstand the maximum He pressure of 8 MPa. Heat transfer coefficients (HTC) obtained form the correlations revealed that required cooling could be achieved by artificially roughened surface towards the plasma-side wall of He cooling channel which helps to keep the RAFMS temperatures below the allowable limit. A 1D analytical and 2D thermal-hydraulic simulation studies using ANSYS has been performed based on the heat load obtained from neutronics calculations to confirm the heat removal and structural integrity under various conditions including ITER transient events. The required helium flow through the cooling channels are evaluated and used to optimize the suitable header design. The detail design of FW thermal-hydraulics, thermostructural analyses, and He flow distribution network will be presented in this paper.

ITER Driver Blanket, European Community design

Fusion Engineering and Design, 1993

Depending on the final decision on the operation time of ITER (International Thermonuclear Experimental Reactor), the Driver Blanket might become a basic component of the machine with the main function of producing a significant fraction (close to 0.8) of the tritium required for the ITER operation, the remaining fraction being available from external supplies.

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.