Breeding Blankets

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David Demange - One of the best experts on this subject based on the ideXlab platform.

  • model and simulation of a vacuum sieve tray for t extraction from liquid pbli Breeding Blankets
    Fusion Engineering and Design, 2016
    Co-Authors: Merlijn Mertens, David Demange, Laetitia Frances
    Abstract:

    Tritium self-sufficiency within a nuclear fusion reactor is necessary to demonstrate nuclear fusion as a viable source of energy. Tritium can be produced within liquid eutectic PbLi but then has to be extracted to be refuelled to the plasma. The vacuum sieve tray (VST) method is based on the extraction of tritium from millimetre-scaled oscillating PbLi droplets falling inside a vacuum chamber. A simulation tool was developed describing the fluid dynamics occurring along the PbLi flow and was used to study the influence of the different geometrical and operational parameters on the VST performance. The simulation predicts that extraction efficiencies over 90% can be easily reached according to theory and previous experimental results. The size of the VST extraction unit for a fusion reactor is estimated based on the findings from our single-nozzle model and assuming no T reabsorption. It is found to be in the feasible range. Nevertheless, two approaches are discussed which may further reduce this size by up to 90%. The simulation tool proved to be an easy and powerful way to analyse and optimise VST set-ups at any scale.

  • tritium extraction technologies and demo requirements
    Fusion Engineering and Design, 2016
    Co-Authors: David Demange, Laetitia Frances, Alessia Santucci, Rodrigo Antunes, O Borisevich, D Rapisarda, M Utili
    Abstract:

    Abstract The conceptual design of the tritium extraction system (TES) for the European DEMO reactor is worked out in parallel for four different Breeding Blankets (BB) retained by EUROfusion. The TES design has to be tackled in an integrated manner optimizing the synergy with the directly interfacing inner fuel cycle, while minimizing the tritium permeation into the coolant. Considering DEMO requirements, it is most likely that only advanced technologies will be suitable for the tritium extraction systems of the BB. This paper overviews the European work programme for R&D on tritium technology for the DEMO BB, summaries the general first outcomes, and details the specific and comprehensive R&D program to study experimentally immature but promising technologies such as vacuum sieve tray or permeator against vacuum for tritium extraction from PbLi, and advanced inorganic membranes and catalytic membrane reactor for tritium extraction from He. These techniques are simple, fully continuous, likely compact with contained energy consumption. Several European Laboratories are joining their efforts to deploy several new experimental setups to accommodate the tests campaigns that will cover small scale experiments with tritium and inactive medium scale tests so as to improve the technology readiness level of these advanced processes.

  • tritium management and anti permeation strategies for three different Breeding blanket options foreseen for the european power plant physics and technology demonstration reactor study
    Fusion Engineering and Design, 2014
    Co-Authors: David Demange, L V Boccaccini, F Franza, A Santucci, S. Tosti, R Wagner
    Abstract:

    Abstract In DT fusion reactors like DEMO, the commonly accepted tritium (T) losses through the steam generator (SG) shall not exceed about 2 mg/d that are more than 5 orders of magnitude lower than the T production rate of about 360 g/d in the Breeding blanket (BB). A very effective mitigation strategy is required balancing the size and efficiency of the processes in the Breeding and cooling loops, and the availability and efficiency of anti-permeation barriers. A numerical study is presented using the T permeation code FUS-TPC that computes all T flows and inventories considering the design and operation of the BB, the SG, and the T systems. Many scenarios are numerically analyzed for three Breeding Blankets concepts – helium cooled pebbles bed (HCPB), helium cooled lithium lead (HCLL), and water cooled lithium lead (WCLL) – varying the T processes throughput and efficiency, and the permeation regimes through the BB and SG to be either surface-limited or diffusion-limited with possible permeation reduction factor. For each BB concept, we discuss workable operation scenarios and suggest specific anti-permeation strategies.

  • Tritium Transport Issues for Helium-Cooled Breeding Blankets
    IEEE Transactions on Plasma Science, 2014
    Co-Authors: F. Franza, David Demange, L. V. Boccaccini, A. Ciampichetti, Massimo Zucchetti
    Abstract:

    Tritium mobility through Breeding blanket (BB) and steam generator heat transfer areas is a crucial aspect for the design of the next generation DEMO fusion power plants. Tritium is generated inside the breeder, dissolves in and permeates through materials, thus leading to a potential hazard for the environment. For this reason, it is important to carry out the tritium migration analysis for a specific DEMO blanket configuration to predict the released amount of tritium during the plant operation. Unfortunately, tritium assessments are often affected by several uncertainties implying very important modeling and parametric issues. In this paper, the main permeation issues are identified and possible solutions are discussed to address the modeling issues and the parametric uncertainties affecting the T migration assessments for the two DEMO helium-cooled BBs: 1) helium-cooled pebble beds and 2) helium-cooled lithium-lead. For these two helium-cooled blanket concepts various tritium migration analyses will be carried out by means of the computational tool FUS-TPC to define proper and feasible tritium mitigation techniques, which are needed to keep the tritium losses lower than the allowable environmental release (i.e., 20 Ci/d).

  • tritium management and safety issues in iter and demo Breeding Blankets
    Fusion Engineering and Design, 2013
    Co-Authors: Beate Bornschein, David Demange, T Pinna
    Abstract:

    Abstract Safe, reliable and efficient tritium management in the breeder blanket faces unique technological challenges. Beside the tritium recovery efficiency in the tritium extraction and coolant purification systems, the tritium tracking accuracy between the inner and outer fuel cycle shall also be demonstrated. Furthermore, it is self-evident that safe handling and confinement of tritium need to be stringently assured to evolve fusion as a reliable technique. The present paper gives an overview of tritium management in breeder Blankets. After a short introduction into the tritium fuel cycle and blanket basics, open tritium issues are discussed, thereby focusing on tritium extraction from blanket, coolant detritiation and tritium analytics and accountancy, necessary for accurate and reliable processing as well as for book-keeping.

L.a. Sedano - One of the best experts on this subject based on the ideXlab platform.

  • Numeric implementation of a nucleation, growth and transport model for helium bubbles in lead–lithium HCLL Breeding blanket channels: Theory and code development
    Fusion Engineering and Design, 2020
    Co-Authors: L. Batet, J. Fradera, E. Mas De Les Valls, L.a. Sedano
    Abstract:

    Large helium (He) production rates in liquid metal Breeding Blankets of a DT fusion reactor might have a significant influence in the system design. Low He solubility together with high local concentrations may create the conditions for He cavitation, which would have an impact in the components performance. The paper states that such a possibility is not remote in a helium cooled lithium–lead Breeding blanket design. A model based on the Classical Nucleation Theory (CNT) has been developed and implemented in order to have a specific tool able to simulate HCLL systems and identify the key parameters and sensitivities. The nucleation and growth model has been implemented in the open source CFD code OpenFOAM so that transport of dissolved atomic He and nucleated He bubbles can be simulated. At the current level of development it is assumed that void fraction is small enough not to affect either the hydrodynamics or the properties of the liquid metal; thus, bubbles can be represented by means of a passive scalar. He growth and transport has been implemented using the mean radius approach in order to save computational time. Limitations and capabilities of the model are shown by means of zero-dimensional simulation and sensitivity analysis under HCLL Breeding unit conditions.Postprint (published version

  • Implementation of two-phase tritium models for helium bubbles in HCLL Breeding blanket modules
    Journal of Nuclear Materials, 2011
    Co-Authors: J. Fradera, L.a. Sedano, E. Mas De Les Valls, L. Batet
    Abstract:

    Tritium self-sufficiency requirement of future DT fusion reactors involves large helium production rates in the Breeding Blankets; this might impact on the conceptual design of diverse fusion power reactor units, such as Liquid Metal (LM) Blankets. Low solubility, long residence-times and high production rates create the conditions for Helium nucleation, which could mean effective T sinks in LM channels. A model for helium nano-bubble formation and tritium conjugate transport phenomena in liquid Pb17.5Li and EUROFER is proposed. In a first approximation, it has been considered that He bubbles can be represented as a passive scalar. The nucleation model is based on the classical theory and includes a simplified bubble growth model. The model captures the interaction of tritium with bubbles and tritium diffusion through walls. Results show the influence of helium cavitation on tritium inventory and the importance of simulating the system walls instead of imposing fixed boundary conditions.

  • Numeric implementation of a nucleation, growth and transport model for helium bubbles in lead–lithium HCLL Breeding blanket channels: Theory and code development
    Fusion Engineering and Design, 2011
    Co-Authors: L. Batet, J. Fradera, E. Mas De Les Valls, L.a. Sedano
    Abstract:

    Large helium (He) production rates in liquid metal Breeding Blankets of a DT fusion reactor might have a significant influence in the system design. Low He solubility together with high local concentrations may create the conditions for He cavitation, which would have an impact in the components performance. The paper states that such a possibility is not remote in a helium cooled lithium–lead Breeding blanket design. A model based on the Classical Nucleation Theory (CNT) has been developed and implemented in order to have a specific tool able to simulate HCLL systems and identify the key parameters and sensitivities. The nucleation and growth model has been implemented in the open source CFD code OpenFOAM so that transport of dissolved atomic He and nucleated He bubbles can be simulated. At the current level of development it is assumed that void fraction is small enough not to affect either the hydrodynamics or the properties of the liquid metal; thus, bubbles can be represented by means of a passive scalar. He growth and transport has been implemented using the mean radius approach in order to save computational time. Limitations and capabilities of the model are shown by means of zero-dimensional simulation and sensitivity analysis under HCLL Breeding unit conditions.

  • numerical analysis of mhd thermofluid flows considering sandwich structures as applied to liquid Breeding Blankets for fusion technology
    8th PAMIR International Conference on Fundamental and Applied MHD, 2011
    Co-Authors: Elisabet Mas De Les Valls Ortiz, Lluis Batet Miracle, Vicente Cesar De Medina Iglesias, Jordi Fradera, L.a. Sedano
    Abstract:

    Lead lithium flowing inside Breeding blanket’s channels in a fusion reactor is subject to a Hartmann number of 104 and a Grashof number 109���12. Since under such conditions buoyancy can be of the same order of magnitude as the electromagnetic force, a deep understanding of the coupled phenomena is required. In the present study, a horizontal channel with dimensions, mean velocity and thermal load taken from HCLL blanket design, and considering a Hartmann number of 3000, is analysed using di erent thermal strategies. The unstable nature of the resulting flow is studied, as well as its influence on relevant thermal parameters.

Laetitia Frances - One of the best experts on this subject based on the ideXlab platform.

  • accuracy evaluation and experimental plan of the multi nozzle vacuum sieve tray facility at the tritium laboratory karlsruhe
    Fusion Engineering and Design, 2019
    Co-Authors: Ester Diazalvarez, Laetitia Frances
    Abstract:

    Abstract Tritium will be produced in Breeding Blankets by neutron bombardment of lithium to ensure the self-sufficiency of fusion power plants. Then, this tritium must be extracted to fuel the plasma. The Vacuum Sieve Tray (VST) technique has been proposed for the tritium extraction system of liquid Blankets in the European DEMO. This technique consists in extracting the tritium dissolved in Pb-16Li by generating droplets, which oscillate while falling in vacuum. The Multi-Nozzle VST (MNVST) setup was assembled at the Tritium Laboratory Karlsruhe (TLK) to study the scalability of the VST technique, as well as to serve as a preliminary deuterium/lead-lithium facility before the construction of a new rig to be operated with tritium. Numerical simulations were discussed with regard to the expected accuracy of measurements to develop an experimental plan. The amount of deuterium extracted depending on the operation conditions was estimated and its distinguishability from one experiment to another was investigated, resulting in the approval of the methodology. As a result, six equilibrium pressures (10, 50, 100, 200, 300, 400 mbar), three Pb-16Li temperatures (350, 400, 450 °C) and two nozzle geometries (1 and 19 nozzles) were selected to be tested during the MNVST experiments.

  • model and simulation of a vacuum sieve tray for t extraction from liquid pbli Breeding Blankets
    Fusion Engineering and Design, 2016
    Co-Authors: Merlijn Mertens, David Demange, Laetitia Frances
    Abstract:

    Tritium self-sufficiency within a nuclear fusion reactor is necessary to demonstrate nuclear fusion as a viable source of energy. Tritium can be produced within liquid eutectic PbLi but then has to be extracted to be refuelled to the plasma. The vacuum sieve tray (VST) method is based on the extraction of tritium from millimetre-scaled oscillating PbLi droplets falling inside a vacuum chamber. A simulation tool was developed describing the fluid dynamics occurring along the PbLi flow and was used to study the influence of the different geometrical and operational parameters on the VST performance. The simulation predicts that extraction efficiencies over 90% can be easily reached according to theory and previous experimental results. The size of the VST extraction unit for a fusion reactor is estimated based on the findings from our single-nozzle model and assuming no T reabsorption. It is found to be in the feasible range. Nevertheless, two approaches are discussed which may further reduce this size by up to 90%. The simulation tool proved to be an easy and powerful way to analyse and optimise VST set-ups at any scale.

  • tritium extraction technologies and demo requirements
    Fusion Engineering and Design, 2016
    Co-Authors: David Demange, Laetitia Frances, Alessia Santucci, Rodrigo Antunes, O Borisevich, D Rapisarda, M Utili
    Abstract:

    Abstract The conceptual design of the tritium extraction system (TES) for the European DEMO reactor is worked out in parallel for four different Breeding Blankets (BB) retained by EUROfusion. The TES design has to be tackled in an integrated manner optimizing the synergy with the directly interfacing inner fuel cycle, while minimizing the tritium permeation into the coolant. Considering DEMO requirements, it is most likely that only advanced technologies will be suitable for the tritium extraction systems of the BB. This paper overviews the European work programme for R&D on tritium technology for the DEMO BB, summaries the general first outcomes, and details the specific and comprehensive R&D program to study experimentally immature but promising technologies such as vacuum sieve tray or permeator against vacuum for tritium extraction from PbLi, and advanced inorganic membranes and catalytic membrane reactor for tritium extraction from He. These techniques are simple, fully continuous, likely compact with contained energy consumption. Several European Laboratories are joining their efforts to deploy several new experimental setups to accommodate the tests campaigns that will cover small scale experiments with tritium and inactive medium scale tests so as to improve the technology readiness level of these advanced processes.

L. Batet - One of the best experts on this subject based on the ideXlab platform.

  • Numeric implementation of a nucleation, growth and transport model for helium bubbles in lead–lithium HCLL Breeding blanket channels: Theory and code development
    Fusion Engineering and Design, 2020
    Co-Authors: L. Batet, J. Fradera, E. Mas De Les Valls, L.a. Sedano
    Abstract:

    Large helium (He) production rates in liquid metal Breeding Blankets of a DT fusion reactor might have a significant influence in the system design. Low He solubility together with high local concentrations may create the conditions for He cavitation, which would have an impact in the components performance. The paper states that such a possibility is not remote in a helium cooled lithium–lead Breeding blanket design. A model based on the Classical Nucleation Theory (CNT) has been developed and implemented in order to have a specific tool able to simulate HCLL systems and identify the key parameters and sensitivities. The nucleation and growth model has been implemented in the open source CFD code OpenFOAM so that transport of dissolved atomic He and nucleated He bubbles can be simulated. At the current level of development it is assumed that void fraction is small enough not to affect either the hydrodynamics or the properties of the liquid metal; thus, bubbles can be represented by means of a passive scalar. He growth and transport has been implemented using the mean radius approach in order to save computational time. Limitations and capabilities of the model are shown by means of zero-dimensional simulation and sensitivity analysis under HCLL Breeding unit conditions.Postprint (published version

  • Implementation of two-phase tritium models for helium bubbles in HCLL Breeding blanket modules
    Journal of Nuclear Materials, 2011
    Co-Authors: J. Fradera, L.a. Sedano, E. Mas De Les Valls, L. Batet
    Abstract:

    Tritium self-sufficiency requirement of future DT fusion reactors involves large helium production rates in the Breeding Blankets; this might impact on the conceptual design of diverse fusion power reactor units, such as Liquid Metal (LM) Blankets. Low solubility, long residence-times and high production rates create the conditions for Helium nucleation, which could mean effective T sinks in LM channels. A model for helium nano-bubble formation and tritium conjugate transport phenomena in liquid Pb17.5Li and EUROFER is proposed. In a first approximation, it has been considered that He bubbles can be represented as a passive scalar. The nucleation model is based on the classical theory and includes a simplified bubble growth model. The model captures the interaction of tritium with bubbles and tritium diffusion through walls. Results show the influence of helium cavitation on tritium inventory and the importance of simulating the system walls instead of imposing fixed boundary conditions.

  • Numeric implementation of a nucleation, growth and transport model for helium bubbles in lead–lithium HCLL Breeding blanket channels: Theory and code development
    Fusion Engineering and Design, 2011
    Co-Authors: L. Batet, J. Fradera, E. Mas De Les Valls, L.a. Sedano
    Abstract:

    Large helium (He) production rates in liquid metal Breeding Blankets of a DT fusion reactor might have a significant influence in the system design. Low He solubility together with high local concentrations may create the conditions for He cavitation, which would have an impact in the components performance. The paper states that such a possibility is not remote in a helium cooled lithium–lead Breeding blanket design. A model based on the Classical Nucleation Theory (CNT) has been developed and implemented in order to have a specific tool able to simulate HCLL systems and identify the key parameters and sensitivities. The nucleation and growth model has been implemented in the open source CFD code OpenFOAM so that transport of dissolved atomic He and nucleated He bubbles can be simulated. At the current level of development it is assumed that void fraction is small enough not to affect either the hydrodynamics or the properties of the liquid metal; thus, bubbles can be represented by means of a passive scalar. He growth and transport has been implemented using the mean radius approach in order to save computational time. Limitations and capabilities of the model are shown by means of zero-dimensional simulation and sensitivity analysis under HCLL Breeding unit conditions.

S. Tosti - One of the best experts on this subject based on the ideXlab platform.

  • Membrane Processes for the Nuclear Fusion Fuel Cycle
    Membranes, 2018
    Co-Authors: S. Tosti, Alfonso Pozio
    Abstract:

    This paper reviews the membrane processes for the nuclear fusion fuel cycle—namely, the treatment of the plasma exhaust gases and the extraction of tritium from the Breeding Blankets. With respect to the traditional processes, the application of membrane reactors to the fusion fuel cycle reduces the tritium inventory and processing time, thus increasing the safety and availability of the system. As an example, self-supported Pd-alloy membrane tubes have been studied for the separation of hydrogen and its isotopes from both gas- and liquid-tritiated streams through water-gas shift and isotopic swamping reactions. Furthermore, this paper describes an innovative membrane system (Membrane Gas–Liquid Contactor) for the extraction of hydrogen isotopes from liquid LiPb Blankets. Porous membranes are exposed to the liquid metal that penetrates the pores without passing through them, then realizing a gas–liquid interface through which the mass transfer of hydrogen isotopes takes place. Compared to the conventional hydrogen isotope extraction processes from LiPb that use the “permeator against vacuum” concept, the proposed process significantly reduces mass-transfer resistance by improving the efficiency of the tritium recovery system.

  • tritium management and anti permeation strategies for three different Breeding blanket options foreseen for the european power plant physics and technology demonstration reactor study
    Fusion Engineering and Design, 2014
    Co-Authors: David Demange, L V Boccaccini, F Franza, A Santucci, S. Tosti, R Wagner
    Abstract:

    Abstract In DT fusion reactors like DEMO, the commonly accepted tritium (T) losses through the steam generator (SG) shall not exceed about 2 mg/d that are more than 5 orders of magnitude lower than the T production rate of about 360 g/d in the Breeding blanket (BB). A very effective mitigation strategy is required balancing the size and efficiency of the processes in the Breeding and cooling loops, and the availability and efficiency of anti-permeation barriers. A numerical study is presented using the T permeation code FUS-TPC that computes all T flows and inventories considering the design and operation of the BB, the SG, and the T systems. Many scenarios are numerically analyzed for three Breeding Blankets concepts – helium cooled pebbles bed (HCPB), helium cooled lithium lead (HCLL), and water cooled lithium lead (WCLL) – varying the T processes throughput and efficiency, and the permeation regimes through the BB and SG to be either surface-limited or diffusion-limited with possible permeation reduction factor. For each BB concept, we discuss workable operation scenarios and suggest specific anti-permeation strategies.

  • Impact of tritium solubility in liquid Pb-17Li on tritium migration in HCLL and WCLL Blankets
    2013 IEEE 25th Symposium on Fusion Engineering (SOFE), 2013
    Co-Authors: A Santucci, F Franza, A. Ciampichetti, D. Demange, S. Tosti
    Abstract:

    The next generation of fusion power plants (DEMO) should rely on a Breeding blanket able to efficiently convert into heat the neutrons kinetics energy, to ensure the tritium self-sufficiency and to adequately shield the Toroidal Field Coils from neutrons and gamma rays. The eutectic lithium-lead alloy is a consolidate liquid blanket material which simultaneously includes the breeder (Li) and the neutron multiplier (Pb). The assessment of the tritium inventory inside the blanket and its environmental release requires the knowledge of the hydrogen isotopes interactions with blanket materials, in particular the hydrogen solubility in lithium-lead which is defined by means of the hydrogen Sievert's constant. Several experiments, aiming to determine the hydrogen isotopes solubility in lithium-lead, have been performed in the past giving values of the temperature-dependent Sieverts' constant, KS, distributed in a wide range (covering about two orders of magnitude on the Arrhenius plot). Starting from a literature review of KS values, this work provides a parametric analysis for the most influencing parameters related to tritium migration in Helium-Cooled and Water-Cooled Lead-Lithium DEMO Breeding Blankets. This analysis has been performed by using the computational code FUS-TPC for several operative scenarios and considering the different KS values. This study demonstrates that the tritium Sieverts' constant in Pb-17Li has a great impact on the assessment of tritium losses, whose value can spread in more than one order of magnitude. Furthermore, the analysis suggests suitable permeation reduction factors to be adopted in the different scenarios as well as the need of addressing new accurate experiments on the solubility constant.