Design of ITER vacuum vessel in-wall shielding (original) (raw)
Related papers
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.
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.
Design progress of the ITER vacuum vessel sectors and port structures
Fusion Engineering and Design, 2007
Recent progress of the ITER vacuum vessel (VV) design is presented. As the ITER construction phase approaches, the VV design has been improved and developed in more detail with the focus on better performance, improved manufacture and reduced cost. Based on achievements of manufacturing studies, design improvement of the typical VV Sector (#1) has been nearly finalized. Design improvement of other sectors is in progress-in particular, of the VV Sectors #2 and #3 which interface with tangential ports for the neutral beam (NB) injection. For all sectors, the concept for the in-wall shielding has progressed and developed in more detail. The design progress of the VV sectors has been accompanied by the progress of the port structures. In particular, design of the NB ports was advanced with the focus on the beam-facing components to handle the heat input of the neutral beams. Structural analyses have been performed to validate all design improvements.
Fusion Engineering and Design, 1995
Radiation transport calculations have been carried out to aid in the design of the reference shielding and breeding blankets proposed for the International Thermonuclear Experimental Reactor (ITER). The results of analyses suggest that both blanket assemblies in combination with the surrounding vacuum vessel provide adequate shielding for the toroidal field coils and reduce the heating and damage to levels commensurate with design guidelines. The induced activation levels and decay heat generation in the breeding blanket-vacuum vessel may qualify the assemblies for disposal as Class C low level waste based on US regulations. The nuclear performance of the shielding around neutral beam injection ducts and in the vicinity of the divertor vacuum pumping ports was also investigated. Preliminary results suggest that the proposed neutral beam injection and divertor port shielding is satisfactory in both cases.
ITER-FEAT vacuum vessel and blanket design features and implications for the R&D programme
Nuclear Fusion, 2001
A tight fitting configuration of the VV to the plasma aids the passive plasma vertical stability, and ferromagnetic material in the VV reduces the TF ripple. The blanket modules are supported directly by the VV. A full-scale VV sector model has provided critical information related to fabrication technology, and the magnitude of welding distortions and achievable tolerances. This R&D validated the fundamental feasibility of the double-wall VV design. The blanket module configuration consists of a shield body to which a separate first wall is mounted. The separate first wall has a facet geometry consisting of multiple flat panels, where 3-D machining will not be required. A configuration with deep slits minimizes the induced eddy currents and loads. The feasibility and the robustness of solid HIP joining was demonstrated in R&D, by manufacturing and testing several small and medium scale mock-ups and finally two prototypes. Remote handling tests and assembly tests of a blanket module have demonstrated the basic feasibility of its installation and removal.
Design progress of the vacuum vessel sectors and ports towards the ITER construction
Fusion Engineering and Design, 2008
Recent progress of the ITER vacuum vessel (VV) design is presented. As the ITER construction phase approaches, the VV design has been improved and developed in more detail with the focus on better performance, improved manufacture and reduced cost. Based on the achievements of manufacturing studies being performed in cooperation with the ITER participant teams (PTs), design improvement of the typical VV sector (#1, see the legend to figure 1 in this article) has been nearly finalized. Design improvement of other sectors is in progress-in particular, of the VV sectors #2 and #3 which interface with tangential ports for the neutral beam (NB) injection. For all sectors, the concept for the in-wall shielding has progressed and developed in more detail. The design progress of the VV sectors has been accompanied by the progress of the port structures. Structural analyses have been performed to validate all design improvements.
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.
Conceptual design of the tokamak radiation shielding for the Tokamak Physics Experiment (TPX)
15th IEEE/NPSS Symposium. Fusion Engineering
The tokamak radiation shielding includes the neutron and The tokamak shielding must be suffi,.-ient to limit gamma shielding around the torus and penetrations required neutron activation of ex-vessel components _Io levels that to (1) limit activation of components outside the shield to allow hands-on maintenance of these components. The levels that permit hands-on maintenance and (2) limit the total expected fluence is 4 x 1022 deuterium-deuterium nuclear heating of the superconducting coils and cold structure. The primary design drivers are space, the 350"C (D-D) neutrons and 8 x 1020 deuterium-t_um (D-T) bakeout temperature, and cost; therefore, different shield neutrons [1]. materials were used for different shield components and The shielding must also be sufficient to limit the iocatiom. The shielding is divided into three areas: (1) torus maximum integrated heat load into the superconducting shielding around the vacuum vessel, (2) duct shielding around coils and cold structure during deuterium operation to < 5 the vacuum pumping ducts and vertical diagnostic ducts, and kW. The peak initial neutron source rate during deuterium O) penetration shielding in and around the radial ports. The operation is 5 x 1016 D-D neutrons/s and 1 x 1015 D-T major shield components include water between the walls of neutrons/s. The shielding system must be capable of the vacuum vessel, lead monosilicate/boron carbide tiles that continuous operation at 150°C and withstand bakeout of the are attached to the exterior of the vacuum vessel, shield plugs vacuum vessel at 350°C for indefinite periods. that fill the openings of the large radial ports, and polyethylene/lead/boron shield blocks for duct shielding. A description of the shielding configuration and the performance CONFIGURATION REQUIREMENTS AND ESSENTIAL FEATURES and operational requirements will be discussed. *Research sponsored by the Office of Fusion Energy, U.S. equivalent to at least 20 cm of water, the equivalent of at Department of Energy, under Contract No. DE-AC05-least 1 cm of boron carbide, and the equivalent of at least 2 84OR21400 with Martin MariettaEnergy Systems, Inc, cm of elemental lead. The penetration shi$1ding must LllS"111161JTlONi_lblll_l_,_iT I,_ UI_LII'dlTI_III
Nuclear Fusion, 2003
During the preparation of the procurement specifications for long lead-time items, several detailed vacuum vessel (VV) design improvements are being pursued, such as elimination of the inboard triangular support, adding a separate interspace between inner and outer shells for independent leak detection of field joints, and revising the VV support system to gain a more comfortable structural performance margin. Improvements to the blanket design are also under investigation, an inter-modular key instead of two prismatic keys and a co-axial inletoutlet cooling connection instead of two parallel pipes. One of the most important achievements in the VV R&D has been demonstration of the necessary assembly tolerances. Further development of cutting, welding and nondestructive tests (NDT) for the VV has been continued, and thermal and hydraulic tests have been performed to simulate the VV cooling conditions. In FW/blanket and divertor, full-scale prototypical mock-ups of the FW panel, the blanket shield block, and the divertor components, have been successfully fabricated. These results make us confident in the validity of our design and give us possibilities of alternate fabrication methods.