The ITER neutral beam front end components integration (original) (raw)
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Fusion Engineering and Design, 2009
The reference design for the ITER neutral beam injectors incorporated a fast shutter to limit tritium migration to the injector vacuum enclosures. In 2005, a need for an 'absolute' isolation valve was identified to facilitate injector maintenance procedures and protect the system from an in-vessel ingress of coolant event (ICE). An outline concept for an all-metal seal valve was developed during 2006, in close cooperation with the Swiss valve manufacturer VAT. During the following year, it became apparent that the length of beamline available for the valve was significantly less than originally envisaged, resulting in a radical revision of the design concept. A casing length of 760 mm has been achieved by means of major changes to the casing structure, plate dimensions, pendulum mechanism and seal actuators. A concept for a seal protection system has been developed to prevent beam line contamination reaching the valve components and to protect the valve plate from surface heating by plasma radiation. The new design concept has been extensively validated by analysis, including a whole-system FE model of the valve.
Spallation Neutron Source linac vacuum seal design and testing
PACS2001. Proceedings of the 2001 Particle Accelerator Conference (Cat. No.01CH37268)
Optimum seal selection with respect to vacuum performance, RF electrical performance, life, cost, reliability, and ease of installation is of primary importance. This report summarizes an investigation of metallic vacuum seals including the C-seal, energized spring seal, Helicoflex copper delta seal, aluminum delta seal, delta seal with limiting ring, and a prototype copper diamond seal. This study was conducted to support the Drift Tubing Linac (DTL) and Coupled Cavity Linac (CCL) design efforts for the Spallation Neutron Source (SNS) project. A series of vacuum leak rate tests using necessary reduced thickness flange designs has been completed. Copper plated stainless steel flanges, necessary for flanged joints intersecting RF cavities, have also been studied and tested. Detailed structural modeling of the CCL bridge coupler seal and flange has also been completed to verify flange structural integrity and appropriate seal compressive loading.
ITER divertor, design issues and research and development
Fusion Engineering and Design, 1999
Over the period of the ITER Engineering Design Activity (EDA) the results from physics experiments, modelling, engineering analyses and R&D, have been brought together to provide a design for an ITER divertor. The design satisfies all necessary requirements for steady state and transient heat flux, nuclear shielding, pumping, tritium inventory, impurity control, armour lifetime, electromagnetic loads, diagnostics, and remote maintenance. The design consists of 60 cassettes each comprising a cassette body onto which the plasma facing components (PFCs) are mounted. Each cassette is supported by toroidal rails which are attached to the vacuum vessel. For the PFCs the final armour choice is carbon-fibre-composite (CfC) for the strike point regions and tungsten in all remaining areas. R&D has demonstrated that CfC monoblocks can routinely withstand heat loads up to 20 MW m − 2 and tungsten armour \ 10 MW m − 2 . Analysis and experiment show that a CfC armour thickness of 20 mm will provide sufficient lifetime for at least 1000 full power pulses. The thickness of the cassette body is sufficient to shield the vacuum vessel, so that, if necessary, rewelding is possible, and also provides sufficient stiffness against electromagnetically generated loads. The cassette design provides efficient and proven remote maintenance which should allow exchange of a complete divertor within 6 months. (R. Tivey) 0920-3796/99/$ -see front matter © 1999 Elsevier Science S.A. All rights reserved. PII: S 0 9 2 0 -3 7 9 6 ( 9 9 ) 0 0 0 4 7 -2 R. Ti6ey et al. / Fusion Engineering and Design 46 (1999) 207-220 208
ITER vacuum vessel design and construction
Fusion Engineering and …, 2010
According to recent design review results, the original reference vacuum vessel (VV) design was selected with a number of modifications including 3D shaping of the outboard inner shell. The VV load conditions were updated based on reviews of the plasma disruption and vertical displacement event (VDE) database. The lower port gussets have been reinforced based on structural analysis results, including non-linear buckling. Design of in-vessel coils for the mitigation of edge localized modes (ELM) and plasma vertical stabilization (VS) has progressed. Design of the in-wall-shielding (IWS) has progressed in details. The detailed layout of ferritic steel plates and borated steel plates is optimized based on the toroidal field ripple analysis. The procurement arrangements (PAs) for the VV including ports and IWS have been prepared or signed. Final design reviews were carried out to check readiness for the PA signature. The procedure for licensing the ITER VV according to the French Order on Nuclear Pressure Equipment (ESPN) has started and conformity assessment is being performed by an Agreed Notified Body (ANB). A VV design description document, VV load specification document, hazard and stress analysis reports and particular material appraisal were submitted according to the guideline and RCC-MR requirements.
Fusion Engineering and Design, 2019
The dome and reflector plate are the parts of divertor plasma facing components (PFCs) of ITER tokamak which are mainly aimed for the removal of heat load of maximum 5 MW/m 2 in steady state condition. The dome is a curved tungsten armoured component and the reflector plate is a straight component. These components have multi-layered joints made of various materials such as tungsten (W), OFHC copper (Cu), copper alloy (CuCrZr) and stainless steel (SS316LN). Joining of such multi-layered joints is known to be problematic due to joining of several dissimilar materials. In this paper, we report the indigenous development of medium size dome and reflector plate via vacuum brazing route for ITER like tokamak application. In order to evaluate the performance of the dome against ITER-like scenarios (maximum heat flux removal of 5MW/m 2), the dome has been successfully tested for 1000 number of steady-state thermal cycles at incident heat fluxes of 3.87 MW/m 2 in the High Heat Flux Test Facility (HHFTF) at IPR. Subsequent testing of additional 200 thermal cycles was also done at incident heat flux of 6 MW/m 2. During the High heat flux (HHF) tests, surface temperature of W tiles reached 640 o C and the beam power was restricted at 6MW/m 2 to limit the temperature below 450 o C at the CuCrZr heat sink. Total 1200 steady-state thermal cycles have been completed. At 6 MW/m 2 , the absorbed heat flux was 4 MW/m 2. Engineering analysis on the HHFT of the dome has been performed using Finite element method (FEM) and Computational Fluid Dynamics (CFD) to simulate and to correlate with the experimental data. Ultrasonic immersion technique-Non destructive testing (NDT) was used to inspect the brazed joint quality of the dome before and after the HHFT. The results of the experimental details, engineering analysis and methodology adopted to fabricate the medium size dome and reflector plate are presented here.
Vacuum seals design and testing for a linear accelerator of the National Spallation Neutron Source
2000
of the University ofCal~omia, theUnited States Government, or any agency thereoJ The views and opinwns ofauthors expressed herein do not necessarily state or rejlect those of The Regents ofthe University ofCall@nia, the United States Government, or any agency thereofi Los Alamos National Ldoratoy strongly supports academicji-eedom and a researcher's right to publish; as an institution, however, the Laboratoy does not endorse the viewpoint ofapublication or guarantee its technical correctness. .2
Design finalization and start of construction of ITER vacuum vessel
Fusion Engineering and Design, 2011
The vacuum vessel (VV) design is being finalized including interface components, such as the support rails and feedthroughs of coils for mitigation of edge localized modes (ELM) and vertical stabilization (VS) of the plasma (ELM/VS coils). It was necessary to make adjustments in the locations of the blanket supports and manifolds to accommodate the design modifications in the ELM/VS coils. The lower port gussets were reinforced to keep a sufficient margin under the increased VV load conditions. The VV support design is being finalized as well, with an emphasis on structure simplification. The design of the inwall shielding (IWS) has progressed, considering assembly and required tolerances. The layout of ferritic steel plates and borated steel plates will be optimized based on on-going toroidal field ripple analysis. The VV instrumentation was defined in detail. Strain gauges, thermocouples, displacement meters and accelerometers shall be installed to monitor the status of the VV in normal and off-normal conditions to confirm all safety functions are performed correctly. The ITER VV design was preliminarily approved, and the VV materials including 316L(N) IG were already qualified by the Agreed Notified Body (ANB) according to the procedure of Nuclear Pressure Equipment Order.
Status of R&D of the Plasma Facing Components for the ITER Divertor “
The paper reports the progress made by the ITER Home Teams in the development of robust carbon and tungsten armoured plasma facing components for the ITER divertor. The activities on the development and study of armour materials, joining technologies, non-destructive evaluation techniques, high heat flux testing of manufactured components and neutron irradiation resistance studies are presented. The results of these activities confirm the feasibility of the main divertor components. Examples of the fruitful collaboration between Parties and future R&D needs are also described.
Development of Remote Handleable Axially Decoupled Radiation Resistant Vacuum Seal
2019
Advanced Rare IsotopE Laboratory (ARIEL) facility is a major expansion of TRIUMF's rare isotope research program. Aiming to triple the production of rare isotopes, ARIEL facility includes the new electron linac driver and two target stations for electron and proton beams. Particularities of ARIEL target stations design define the requirements for vacuum interfaces with both primary electron and proton beamlines and rare-isotope beamlines. None of the existing products fully met the requirements, driving the development of custom vacuum interfaces. The design of new vacuum seals is driven both by unique design specifications (limited amount of allowed axial forces, extreme radiation resistance, remote handleability and high repeatability) as well as limitations of the proposed design of beamline infrastructure in the target hall (limited available space and the choice of materials for certain components). This paper discusses preliminary results of the vacuum seal development and...