The water-cooled lithium-lead breeding blanket is in the pre-conceptual design phase. It is a candidate option for European DEMO nuclear fusion reactor. This breeding blanket concept relies on the ...liquid lithium-lead as breeder-multiplier, pressurized water as coolant and EUROFER as structural material. Current design is based on DEMO 2017 specifications. Two separate water systems are in charge of cooling the first wall and the breeding zone: thermo-dynamic cycle is 295–328 °C at 15.5 MPa. The breeder enters and exits from the breeding zone at 330 °C. Cornerstones of the design are the single module segment approach and the water manifold between the breeding blanket box and the back supporting structure. This plate with a thickness of 100 mm supports the breeding blanket and is attached to the vacuum vessel. It is in charge to withstand the loads due to normal operation and selected postulated initiating events. Rationale and progresses of the design are presented and substantiated by engineering evaluations and analyses. Water and lithium lead manifolds are designed and integrated with the two consistent primary heat transport systems, based on a reliable pressurized water reactor operating experience, and six lithium lead systems. Open issues, areas of research and development needs are finally pointed out.
The scope of the EUFOfusion Work Package Breeding Blanket is to develop a blanket concept for the EU DEMO reactor; this includes the blanket segments inside the Vacuum Vessel and the related Tritium ...Extraction/Removal Systems. In the Pre-Concept Design (PCD) Phase, two concepts have been selected as candidates; a solid and a liquid breeder blanket cooled with helium and water, respectively. The design of these two blanket systems has been adapted to the DEMO plant design developed in the PCD Phase and performances assessed. A large R&D programme has been implemented with the scope to evaluate different technologies for these blankets; including the development of breeders, tritium extraction and cooling technologies, and the manufacturing of the blanket system. A major milestone in the subsequent Concept Design Phase is the final selection of the blanket concept for DEMO.
Summary
The water‐cooled lithium–lead breeding blanket is a candidate option for the European Demonstration Power Plant (DEMO) nuclear fusion reactor. This breeding blanket concept relies on the ...liquid lithium–lead as breeder–multiplier, pressurized water as coolant, and EUROFER as structural material. The current design is based on DEMO 2015 specifications and represents the follow‐up of the design developed in 2015. The single‐module‐segment approach is employed. This is constituted by a basic geometry repeated along the poloidal direction. The power is removed by means of radial–toroidal (i.e., horizontal) water cooling tubes in the breeding zone. The lithium–lead flows in a radial–poloidal direction. On the back of the segment, a 100‐mm‐thick plate is in charge of withstanding the loads due to normal operation and selected postulated initiating events. Water and lithium–lead manifolds are designed and integrated with a consistent primary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium–lead system. Rationale and features of the single‐module‐segment water‐cooled lithium–lead breeding blanket design are discussed and supported by thermo‐mechanic, thermo‐hydraulic, and neutronic analyses. Preliminary integration with the primary heat transfer system, the energy storage system, and the balance of plant is briefly discussed. Open issues, areas of research, and development needs are finally pointed out. @EUROfusion Consortium*, 2017. *For more details see http://www.euro‐fusionscipub.org/disclaimer‐copyright
The paper concerns the new design of the water‐cooled lithium–lead breeding blanket, which is one of the candidate options for the European Demonstration Power Plant fusion reactor. The new design consists of a single‐module blanket segment concept and presents horizontal (i.e., radial–toroidal) water‐cooling tubes in the breeding zone, and lithium–lead flowing in the radial–poloidal direction. The paper presents the rationale and features of the water‐cooled lithium–lead breeding blanket design, supported by thermo‐mechanic, thermo‐hydraulic, magneto‐hydrodynamic, and neutronic analyses.
The in-box LOCA (Loss of Coolant Accident) represents a major safety concern to be addressed in the design of the WCLL-BB (Water-Cooled Lead-Lithium Breeding Blanket) that will be tested in the Test ...Blanket Module of the ITER reactor. In this accident, a tube rupture in the coolant circuit would cause a potential disruptive exothermic chemical reaction between the water (the primary coolant) and the Lead-Lithium. This paper presents the development of numerical tools for the analysis of the in-box LOCA consequences in the breeding unit of the blanket.
In this work a simplified computational model of the breeding unit was created by SIMMER-IV which includes the lead lithium and water systems and the internal solid structures. The simulation is set up at the foreseen nominal working conditions of the units and involves a small rupture of the cooling tube, with the consequent propagation of high pressure waves on the breeder side and the fast evolution of the chemical reaction. It is shown here that the code is capable of correctly predicting both these events, from a qualitative point of view, thus providing a highly valuable insight of the whole scenario. These preliminary results will provide a crucial base for further development of this computational tool and the simulation of the whole Test Blanket Module.
Water-cooled lithium-lead breeding blanket is considered a candidate option for European DEMO nuclear fusion reactor. ENEA and the linked third parties have proposed and are developing a multi-module ...blanket segment concept based on DEMO 2015 specifications. The layout of the module is based on horizontal (i.e. radial-toroidal) water-cooling tubes in the breeding zone, and on lithium lead flowing in radial-poloidal direction. This design choice is driven by the rationale to have a modular design, where a basic geometry is repeated along the poloidal direction. The modules are connected with a back supporting structure, designed to withstand thermal and mechanical loads due to normal operation and selected postulated accidents. Water and lithium lead manifolds are designed and integrated with a consistent primary heat transport system, based on a reliable pressurized water reactor operating experience, and the lithium lead system. Rationale and features of current status of water-cooled lithium-lead breeding blanket design are discussed and supported by thermo-mechanics, thermo-hydraulics and neutronics analyses. Open issues and areas of research and development needs are finally pointed out.
The Water-Cooled Lithium Lead breeding blanket concept foresees the eutectic lithium-lead (Pb-15.7Li) alloy being cooled by pressurized sub-cooled water (temperature 295–328 °C; pressure 15.5 MPa) ...flowing in double wall tubes. Therefore, the interaction between the Pb-15.7Li and water (e.g. tube rupture) represents one of the main safety concerns for the design and safety analysis. Available LIFUS5/Mod2 experimental data are employed to assess the performances of thermal-hydraulic and thermo-mechanic codes. Thermal-hydraulic simulations, by SIMMER-III code, are focused on the prediction of the thermodynamic interaction among the fluids. ABAQUS Finite Element code, used for the design activities, is adotped to perform the thermo-mechanic simulations, calculating the stress and strain fields of LIFUS5/Mod2 main vessel during the experiments. Code results are compared with the experimental data and the outcomes from the analyses are discussed, in order to derive conclusions on the code assessment.
The European liquid Test Blanket Module (TBM) concept named Water Cooled Lithium Lead makes use of the liquid metal alloy PbLi as tritium breeder/neutron multiplier and water as coolant. A dedicated ...PbLi ancillary system feeds the detritiated liquid metal alloy to the TBM and returns the PbLi enriched with tritium. The object of this study is developing a numerical model able to simulate the thermo-hydraulic behavior of the PbLi ancillary system under both normal operation and accidental scenario, with the aim to investigate the propagation of pressure waves generated by an in-box LOCA. The accident consists of a tube rupture of the coolant circuit inside the TBM, leading to the interaction between PbLi and water. As a consequence, a pressure increase occurs inside the breeder zone. The accident might affect the ancillary systems due to the propagation of the pressure waves generated inside the TBM. The PbLi loop is modelled by means of the thermo-hydraulics system code RELAP5/Mod3.3 and an instantaneous pressure peak is applied inside the control volume simulating the TBM soon after reaching the steady state conditions inside the whole loop. The propagation of the pressure wave inside the circuit and the hydraulic behavior of the individual components is studied. The present paper represents a first step towards the investigation of the PbLi system and its safety features with the aim to support the R&D activities carried out in the framework of the ITER project.
•The preliminary Pb15.7Li Loop layout was shown in the paper.•The Magneto Hydro Dynamic effect induced in the flowing Pb-16Li inside BB and the pressure drops generated were analysed.•The developed ...anti-permeation and corrosion barriers inside the BB where shown.•The code developed to analyse the In-box Loca inside WCLL BB was introduced.•Preliminary integration of Pb-16Li loops inside the tokamak building was shown.
Three of the four breeder blanket concepts currently under investigation for the European DEMO Reactor use the eutectic Pb-16Li as breeder material. Those are the Helium Cooled Lithium Lead (HCLL), Water Cooled Lithium Lead (WCLL) and Dual Coolant Lithium Lead (DCLL) blankets. Moreover, the WCLL is one of the blanket concepts that will be qualified in the ITER reactor, therefore the development and design of lead lithium loops and auxiliary systems is essential. The main functional requirements that Pb-16Li systems have to fulfill are:
•to circulate the liquid Pb-16Li through the blanket and ancillaries;•to extract the tritium produced inside the breeder modules from Pb-16Li;•to control Pb-16Li chemistry and to remove accumulated impurities;
The present work aims to describe the activities performed in order to achieve the following objectives: i) design and integration of the Pb-16Li loops inside the tokamak building, ii) development and characterization of antipermeation and anticorrosion coatings on structures in contact with Pb-16Li, iii) development and design of an activation products removal system, iv) design of a chemistry control system for Pb-16Li loops, v) performing magnetohydrodynamic analyses taking into account the impact on heat transfer and tritium transport in breeding blankets and performing safety analyses of water/Pb-16Li interaction due to LOCA inside the WCLL blanket.
•Fusion.•Experimental facility.•ENEA Brasimone R.C.
In the European and international framework, ENEA coordinates the Italian fusion program, supported by linked third parties as universities, ...research institutes and industries. In this context, the Experimental Engineering Division (FSN-ING) is involved in experimental and numerical activities related to Breeding Blanket (BB) and Divertor technologies development.
This paper describes the scientific works presently ongoing at Brasimone R.C. enveloped in LLE, lithium, helium, tritium and pressurized water technologies, characterization of structural materials, analysis of materials corrosion rate and development and qualification of anti-permeation/corrosion barrier. The experimental activities conducted for the investigation of safety-relevant scenarios as In-Box LOCA (LLE-water interaction in the WCLL or shock waves propagation generated by helium injection in LLE in the HCLL/DCLL-BBs) are here reported.
Finally, new activities have been planned to support the Divertor Tokamak Test divertor characterization, the large-scale LLE-water interaction and a LLE components validation in relevant scale for WCLL-BB.
PDF
