Search

Abstract: A concept for a shaped first wall for ITER is presented. While keeping most features of the 2004 FDR wall (modules segmentation, plasma facing components technologies, plasma facing material), this concept provides protection of the lateral faces of the first wall panels against the intense parallel heat flux coming from the plasma. Excessive beryllium temperatures at the panel edges are avoided during regular operation. The intense heat flux at the top of the vessel is accounted for and protection is provided against the shine thru heat flux. Start-up and ramp down using the wall as a limiter is possible for up to 7.5MW, both inboard and outboard. This is rendered possible by the use of 5MW/m2 technology panels for 40% of the panels. [ABSTRACT FROM AUTHOR]

As a consequence of its position and functions, the ITER blanket system will be subjected to significant heat loads under nominal reference conditions. Therefore, the design of its cooling system is particularly demanding. Coolant water is distributed individually to the 440 blanket modules (BMs) through manifold piping, which makes it a highly parallelized system. The mass flow rate distribution is finely tuned to meet all operation constraints: adequate margin to burn out in the plasma facing components, even distribution of water flow among the so-called plasma-facing “fingers” of the Blanket First Wall panels, high enough water flow rate to avoid excessive water temperature in the outlet pipes, maximum allowable water velocity lower than 7 m/s in manifold pipes. Furthermore the overall pressure drop and flow rate in each BM shall be within the fixed specified design limit to avoid an unduly unbalance of cooling among the 440 modules. Analyses have to be carried out following a computational fluid-dynamic (CFD) approach based on the finite volume method and adopting a CFD commercial code to assess the thermal-hydraulic behaviour of each single circuit of the ITER blanket cooling system. This paper describes the code benchmarking needed to determine the best method to get reliable and timely results. Since experimental tests are available in ITER Organization on full scale prototypes of Shield Blocks #08 and #14, CFD analyses have been performed to investigate their fluid-dynamic behaviour under steady state conditions and compare the numerical and experimental results. Results obtained are presented and critically discussed. [ABSTRACT FROM AUTHOR]

• Successful high heat flux test results of ITER First Wall mock-ups after neutron irradiation • High heat flux testing of actively cooled mock-ups manufactured with the relevant technologies in relevant geometry to the ITER First Wall, by the suppliers involved in the series production for ITER • High heat flux testing of neutron-irradiated beryllium flat tiles with hypervapotron heat sink • Details of high heat flux test by electron beam facility, JUDITH facility and details of neutron irradiation in the fission reactor, HFR Eight mock-ups imitating the ITER Enhanced Heat Flux First Wall panel were subjected to thermal fatigue test after neutron irradiation at two dose levels, ≈0.1 dpa and ≈0.5 dpa. All the mock-ups successfully withstood repeated thermal fatigue loads up to 4.7 MW/m2. The test results confirmed the performance of mock-ups in the tested condition, specifically the design and manufacturing technologies of the suppliers of two domestic agencies in charge of the series production. [ABSTRACT FROM AUTHOR]

This paper describes the main activities carried out in the frame of EU-DA prequalification for the supply of Normal Heat Flux (NHF) First Wall (FW) panels to ITER. A key part of these activities is the manufacturing development, the fabrication and the factory acceptance tests of a reduced scale FW prototype (Semi-Prototype (SP)) of the NHF design. The SP has a dimension of 221 mm × 665 mm, corresponding to about 1/6 of a full-scale panel, with six full-scale “fingers” and bearing a total of 84 beryllium tiles. It has been manufactured by the AREVA Company in France. The manufacturing process has made extensive use of Hot Isostatic Pressing, which was developed over more than a decade during the ITER Engineering Design Activity phase. The main manufacturing steps for the Semi-Prototype are recalled, with a summary of the lessons learned and the implications with regard to the design and manufacturing of the full-scale prototype and of the series fabrication of the EU-DA share of the ITER first wall (215 NHF panels). The fabricated SP is then tested under High Heat Flux (HHF) in the dedicated test facility of JUDITH-II in Forschungszentrum Jülich, Germany. The objective of the HHF testing is the demonstration of achieving the requested performance under thermal fatigue. The test protocol and facility qualification are presented and the behaviour of the fingers under the 7500 cycles at 2 MW/m 2 is described in detail. [ABSTRACT FROM AUTHOR]

Abstract: The ITER in-vessel components have experienced a major redesign since the ITER Design Review of 2007. A set of in-vessel vertical stabilization (VS) coils and a set of in-vessel Edge Localized Mode (ELM) control coils have been implemented. The blanket system has been redesigned to include first wall (FW) shaping, to upgrade the FW heat removal capability and to allow for an “in situ” replacement. The blanket manifold system has been redesigned to improve leak detection and localisation. The introduction of a new set of in-vessel coils and the design evolution of the blanket system while the ITER project was entering the procurement phase have proven to be a major engineering challenge. This paper describes the status of the redesign of the in-vessel components and the associated integration issues. [Copyright &y& Elsevier]

Highlights: [•] Detailed design development plan for the ITER tungsten divertor. [•] Latest status of the ITER tungsten divertor design. [•] Brief overview of qualification program for the ITER tungsten divertor and status of R&D activity. [Copyright &y& Elsevier]

Abstract: The ITER plasma-facing components directly face the thermonuclear plasma and include the divertor, the blanket and the test blanket modules with their corresponding frames. The divertor is located at the bottom of the plasma chamber and is aimed at exhausting the major part of the plasma thermal power (including alpha power) and at minimising the helium and impurity content in the plasma. The blanket system provides a physical boundary for the plasma transients and contributes to the thermal and nuclear shielding of the vacuum vessel and external machine components. It consists of modular shielding elements known as blanket modules which are attached to the vacuum vessel. Each blanket module consists of two major components: a plasma-facing first wall panel and a shield block. The test blanket modules are mock-ups of DEMO breeding blankets. There are three ITER equatorial ports devoted to test blanket modules, each of them providing for the allocation of two breeding modules inserted in a steel frame and in front of a shield block. [ABSTRACT FROM AUTHOR]