Search results for "first wall"

showing 6 items of 6 documents

Thermo-mechanical analyses and ways of optimization of the helium cooled DEMO First Wall under RCC-MRx rules

2017

Abstract The EUROfusion Consortium develops a design of a fusion power demonstrator plant (DEMO) in the framework of the European “Horizon 2020” innovation and research program. One of the key components in the fusion reactor is the Breeding Blanket (BB) surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. Among the 4 candidates for the DEMO BB, 2 of them use helium as coolant (HCPB, HCLL), and another one (DCLL) uses helium to cool down the First Wall (FW) only. Due to uncertainties regarding the plasma Heat Flux (HF) load the DEMO BB integrated FW will have to cope with, a set of sensitive thermal and stress analys…

Materials scienceRCC-MRxNuclear engineeringchemistry.chemical_elementBlanket01 natural sciences7. Clean energy010305 fluids & plasmasStress (mechanics)[SPI]Engineering Sciences [physics]Materials Science(all)0103 physical sciencesGeneral Materials ScienceCast3M010306 general physicsDEMOHeliumSettore ING-IND/19 - Impianti NucleariCivil and Structural EngineeringSteady stateBreeding BlanketMechanical EngineeringThermo-mechanicFusion powerCoolantFirst WallchemistryCreepHeat fluxNuclear Energy and EngineeringHCLLDEMO; Breeding; Blanket; HCLL; RCC-MRx; Thermo-mechanics; Cast3M; First WallMaterials Science (all)
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Novel method for determination of tritium depth profiles in metallic samples

2019

Tritium accumulation in fusion reactor materials is considered a serious radiological issue, therefore a lot of effort has been concentrated on the development of radiometric techniques. A novel method, based on gradual dissolution, for the determination of the total tritium content and its depth profiles in metallic samples is demonstrated. This method allows for the measurement of tritium in metallic samples after their exposure to a hydrogen and tritium mixture, tritium containing plasma or after irradiation with neutrons resulting in tritium formation. In this method, successive layers of metal are removed using an appropriate etching agent in the controlled regime and the amount of evo…

inorganic chemicalsfusionNuclear and High Energy PhysicsMaterials scienceNuclear engineeringchemistry.chemical_elementheliumBlanket114 Physical sciences01 natural sciences010305 fluids & plasmasblanketMetalirradiated berylliumjet0103 physical sciencespolycyclic compounds010306 general physicsHeliumbreeding blanketJet (fluid)Fusiontritiumbehaviororganic chemicalshydrogen diffusiontemperatureiter-like-wallFusion powerfirst wallberylliumCondensed Matter Physicschemistryvisual_arttransportcardiovascular systemvisual_art.visual_art_mediumdepth profileTritiumBerylliumNuclear Fusion
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Draining and drying process development of the Tokamak Cooling Water System of ITER

2016

Abstract The ITER Organization (IO) developed a thermal-hydraulic (TH) model of the complex first wall and blanket (FW/BLK) cooling channels to determine gas flow rate and pressure required to effectively blow out the water in the FW/BLK. In addition, US ITER conducted experiments for selected geometries of FW/BLK flow channels to predict the blowout parameters. The analysis indicates that as low as 2 MPa of pressure difference over the blanket modules will ensure substantial evacuation of the water in blankets with just a few percent remaining in the blanket flow channels. A limited validation study indicates that the analysis yields less conservative results to compare against data collec…

Validation studyTokamakProcess developmentMechanical EngineeringNuclear engineeringFlow (psychology)First Wall Blanket Draining Process Drying ProcessBlanketBlow out01 natural sciences010305 fluids & plasmasVolumetric flow ratelaw.inventionNuclear Energy and Engineeringlaw0103 physical sciencesWater coolingEnvironmental scienceGeneral Materials Science010306 general physicsSettore ING-IND/19 - Impianti NucleariCivil and Structural EngineeringFusion Engineering and Design
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On the optimization of the first wall of the DEMO water-cooled lithium lead outboard breeding blanket equatorial module

2016

Abstract Within the framework of EUROfusion R&D activities a research campaign has been carried out at the University of Palermo in order to investigate the thermo-mechanical performances of the DEMO water-cooled lithium lead (WCLL) breeding blanket first wall (FW). The research campaign has been mainly focused on the optimization of the FW geometric configuration in order to maximize the heat flux it may safely withstand fulfilling all the thermal, hydraulic and mechanical requirements foreseen by safety codes. Attention has been focused on the FW flat concept endowed with square cooling channels and the potential influence of its four main geometrical parameters on its thermo-mechanical p…

Steady stateComputer scienceMechanical EngineeringNuclear engineeringchemistry.chemical_elementBlanket01 natural sciences7. Clean energyFinite element methodSquare (algebra)010305 fluids & plasmasDEMO reactor WCLL blanket First wallNuclear Energy and EngineeringHeat fluxchemistry0103 physical sciencesThermalGeneral Materials ScienceLithium010306 general physicsLead (electronics)Settore ING-IND/19 - Impianti NucleariCivil and Structural Engineering
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Numerical thermo-mechanical analysis of the DEMO Water-Cooled Lithium Lead breeding blanket conceptual design

First WallBreeding BlanketThermo-mechanical analysiEUROfusionDEMOSettore ING-IND/19 - Impianti NucleariEUROfusion; DEMO; Thermo-mechanical analysis; Breeding Blanket; First Wall
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Optimization of the first wall for the DEMO water cooled lithium lead blanket

2015

The maximum heat load capacity of a DEMO First Wall (FW) of reasonable cost may impact the decision of the implementation of limiters in DEMO. An estimate of the engineering limit of the FW heat load capacity is an essential input for this decision. This paper describes the work performed to optimize the FW of the Water Cooled Lithium-Lead (WCLL) blanket concept for DEMO fusion reactor in order to increase its maximum heat load capacity. The optimization is based on the use of water at typical Pressurised Water Reactors conditions as coolant. The present WCLL FW with a waved plasma-faced surface and with circular channels was studied and the heat load limit has been predicted with FEM analy…

Work (thermodynamics)DesignMaterials scienceMechanical EngineeringNuclear engineeringDesign;DEMO;Blanket;WCLL;First WallFusion powerBlanketFinite element methodWCLLCoolantDesign for manufacturability[SPI]Engineering Sciences [physics]First WallDEMO Blanket WCLL Design First WallNuclear Energy and EngineeringThermalLimiterBlanketGeneral Materials ScienceDEMOSettore ING-IND/19 - Impianti NucleariCivil and Structural EngineeringFusion Engineering and Design
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