Concept design studies on the ITER HNB Duct Liner (original) (raw)
Related papers
Thermo-hydraulic design verification of the neutral beam liner for the ITER vacuum vessel
Fusion Engineering and Design, 2011
The ITER vacuum vessel has upper, equatorial and lower port structures used for equipment installation, utility feedthroughs, vacuum pumping and access inside the vessel for maintenance. A neutral beam (NB) port of equatorial ports provides a path of neutral beam for plasma heating and current drive. An internal duct liner is built in the NB ports, and copper alloy panels are placed in the top and bottom of the liner to protect duct surface from high-power heat loads. Global NB liner models for the upper panel and the lower panel have been developed, and flow field and conjugate heat transfer analyses have been performed to find out pressure drop and heat transfer characteristics of the liner. Heat loads such as NB power, volumetric heating and surface heat flux are applied in the analyses for beam steering and misalignment conditions. For the upper panel, it is found that unbalanced flow distribution occurs in each flow path, and this causes poor heat transfer performance as well. In order to improve flow distribution and to reduce pressure losses, hydraulic analyses for modified cooling path schemes have been carried out, and design recommendations are made based on the analysis results. For the lower panel, local flow distributions and pressure drop values at each header and branch of the tube are obtained by applying design coolant flow rate. Together with the coolant flow field, temperature and heat transfer coefficient distributions are also acquired from the analyses. Based on the analysis results, it is concluded that the lower panel for the NB liner is relatively well designed even though the given heat loads are very severe.
Fusion Engineering and Design, 2008
This paper describes the analyses performed to investigate and validate the proposed design for the updated JET MKIII duct side liner, which will replace the present inertial cooled one in the frame of the EP2 neutral beam enhancement project. The thermal-hydraulic and thermo-mechanical performance of a duct liner's generic module, under various loading scenarios has been assessed. Due to difference in scale between a generic liner module length and the relevant load bearing section thickness (∼1.2 m against 4 mm) two different scale FE models have been assessed, the first ones to evaluate the overall reactions and displacements and the others to calculate concentrated stresses in the most loaded sections. Conformity to ITER design criteria has been verified for both monotonic and cyclic loads. The effects of fatigue have been considered and an operational life of 8.5 years is predicted for the liner.
Beam duct for the 1 MW neutral beam heating injector on TCV
Fusion Engineering and Design, 2017
The Tokamak à Configuration Variable (TCV) has been recently equipped with a 1 MW neutral beam heating (NBH) injector [1]. Two new stainless steel ports with rectangular aperture of 170x220mm have been manufactured and installed for this purpose. The NBH injector is connected to one of them via a stainless steel port extension. The port and its extension together form the beam duct between the TCV vacuum vessel (VV) and the NBH injector. A preliminary thermal analysis of the beam duct showed no expectation of thermal events such as overheating. Indeed, although the beam power density near the internal faces of the beam duct reaches ~10 MW/m 2 , the very grazing incidence angle between the beam axis and the wall was expected to lead to a maximal effective heat flux of ~350 kW/m 2 for a maximal duration of 2 seconds, resulting in acceptable temperature rise. As a result, the beam duct did not include any provision for cooling. However, early in 2016 the commissioning of the NBH injector showed high overheating of the port extension, resulting in local melting and ultimately vacuum leak. This paper describes the design and analysis of an actively cooled beam duct and the status of the beam duct.
The high heat flux components for ITER neutral beam system
Fusion Engineering and Design, 2000
All the alternatives considered for the reduced technical objectives/reduced cost international thermonuclear experimental reactor (RTO/RC ITER) foresee, as one of the methods for additional heating and current drive, the use of neutral injection. Two atomic deuterium beams, with 1 MeV energy and each with at least 16.7 MW power, are the reference for the engineering design of the system. The design of NB system developed for the ITER final design report (FDR) [1] is being modified to comply with the reduced size of the machine and to incorporate, if possible, some design improvements and simplifications. The analyses of overall power distribution and the maximum power densities on each of the beamline components have been updated. To obtain 16.7 MW neutral beam power, 35 MW ion beam power at the exit of the accelerator is necessary. Therefore, about 18 MW are deposited on the beamline components. Moreover, during the commissioning of the injectors, and the high voltage conditioning of the beam source, the full beam power is dumped and measured on a calorimeter. The beamline requires actively cooled high heat flux (HHF) components to exhaust high power density (up to 15 MW/m 2) in steady state conditions. The operative life of this components (30 000 beam-on/beam off cycles, without replacement) requires, primarily, the verification of thermal fatigue safety margins. This paper describes the most relevant aspects of the mechanical design of the HHF components, focusing on the thermal and mechanical verifications.
Fusion Engineering and Design, 2010
Temperature control of the high heat flux (HHF) components in the neutral beam injection (NBI) system in the Experimental Advanced Superconducting Tokamak (EAST) is a very important issue for realizing a high-parameter steady state condition in the tokamak. This paper presents a 3-D computational fluid dynamics analysis of a cooling system of a residual ion dump (RID) composed of hypervapotron structures, which has been previously used in the Mega-Amp Spherical Tokamak (MAST). The results obtained by the proposed method coincide well with experimental values, which validate this method. This study provides a powerful tool for protecting the HHF components in the NBI system from overheating and sustaining irreparable damage. Moreover, this study provides a potentially useful method for optimizing the structural design of HHF components in the NBI system.
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.
Fusion Engineering and Design, 2008
The International Fusion Materials Irradiation Facility (IFMIF) is devised to contribute experimental evidence to an irradiated material properties database for candidate materials exposed to irradiation spectra and doses relevant for future fusion power reactors. Due to neutron fluxes generated by high-energy deuterons reacting in a liquid lithium target, damage rates of 20-50 displacements per atom in one full power year can be achieved in steel specimens inside a volume of approximately 0.5 L. The design of the high flux test module developed at the Forschungszentrum Karslruhe (FZK) allows for maximizing the space available in the high flux neutron field for material irradiation, while at the same time allowing precise adherence of the irradiation temperature of the specimen stacks. Since enhancement of the neutron irradiation requires placement of the specimens as close as possible to the neutron source, the design proposes thin container structures (obeying mechanical constraints) and flat coolant channels between the rigs. A helium gas flow is designated to remove the heat from the rigs to keep the required irradiation temperature, which may be chosen between 250 and 650 • C. As a result of the thin container walls and the small channel dimensions, the helium cooling gas flow is characterized by low pressure, transitional Reynolds numbers and intermediate Mach numbers.
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.
Design update, thermal and fluid dynamic analyses of the EU-HCPB TBM in vertical arrangement
Fusion Engineering and Design, 2009
In the frame of the activities of the EU Breeder Blanket Programme and of the Test Blanket Working Group of ITER, the Helium Cooled Pebble Bed Test Blanket Module (HCPB TBM) is developed in Forschungszentrum Karlsruhe (FZK) to investigate DEMO relevant concepts for blanket modules. The three main functions of a blanket module (removing heat, breeding tritium and shielding sensitive components from radiation) will be tested in ITER using a series of four TBMs, which are irradiated successively during different test campaigns. Each HCPB TBM will be installed, with a vertical orientation, into the vacuum vessel connected to one equatorial port. As the studies performed up to 2006 in FZK concerned a horizontal orientation of the HCPB TBM, a global review of the design is necessary to match with the new ITER specifications. A preliminary version of the new vertical design is proposed extrapolating the neutronic analysis performed for the horizontal HCPB TBM. An overview of the new HCPB TBM vertical designs, as well as the preliminary thermal and fluid dynamic analyses performed for the validation of the design, are presented in this paper. A critical review of the results obtained allows us, in the conclusion, to prepare a plan for the future detailed analyses of the vertical HCPB TBM.