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One of the main challenges to be addressed to achieve a reliable electricity production from the EU DEMO reactor is the realization of a closed fuel cycle, for which a suitable Tritium Extraction and Removal System (TERS) is required. One of the possible technologies identified for the EU DEMO TERS is the Permeator Against Vacuum (PAV): the tritium dissolved in the liquid PbLi flowing within several parallel channels will permeate towards the vacuum pumped on the other side of the channel wall (the membrane).A recently-developed model of the tritium permeation across the membrane in the PAV, involving both transport phenomena in the wall and surface processes, was already used to size the EU DEMO PAV. However, besides the component itself, it is important to properly define the interfaces of the PAV in the TERS, and of the TERS in the entire PbLi and tritium loops. The model of such a complex system is therefore implemented here in the Modelica object-oriented language used by system-level tool GETTHEM, that already includes a model of the PbLi loop. The resulting, lumped-parameter component will be able to capture the thermal-hydraulic behaviour of the PbLi, to model the tritium transport in the fluid and to estimate the tritium permeated flux supplied to the tritium processing. Such a model is tested here on a sub-scale circuit to demonstrate its capability to simulate the operation of the EU DEMO TERS using the GETTHEM code.As the physical parameters of the model are subject to a large uncertainty, an uncertainty propagation analysis is also performed, to have a preliminary quantification of the impact of such uncertainties on the model output and, therefore, on the TERS efficiency, and to drive further investigations of these physical properties. In particular, results show how the uncertainty on the solubility constant of hydrogen in PbLi represents the dominant contribution on the total variance, highlighting the need for a better accuracy of such parameter.

The Permeator Against Vacuum (PAV) is one of the proposed technologies for the Tritium Extraction System of the WCLL BB (Water-Cooled Lithium-Lead Breeding Blanket) of the EU DEMO reactor. In this paper, the manufacturing of the first PAV mock-up with a niobium membrane with a cylindrical configuration is presented. This work aimed to demonstrate the possibility of manufacturing a relevant-size PAV to be later tested in the TRIEX-II facility. The adopted prototypical solutions are described in detail, starting with the methodology developed to join the Nb tubes with a 10CrMo9-10 (A182 F22) plate. Dedicated manufacturing and welding procedures, based on vacuum brazing with a nickel-based brazing alloy, were developed to solve the problem. This new kind of brazing was first analyzed to check the morphology of the joint and then tested to check its capability to withstand the TRIEX-II operative conditions. In parallel, the compatibility with a lithium-lead environment was analyzed by exposing samples of niobium and 10CrMo9-10 (A335 P22) to a flow of the eutectic alloy at 500 °C up to 4000 h. Finally, the PAV mock-up was installed in the TRIEX-II facility. [ABSTRACT FROM AUTHOR]

Tritium extraction is one of the key open issues toward the development of the WCLL BB (Water-Cooled Lithium–Lead Breeding Blanket) of EU DEMO reactors, and different technologies have been proposed to address it. Among them, the Permeator Against Vacuum (PAV) has promising features, but it has never been tested in a relevant environment. This work presents the first experimental results ever obtained for a PAV mock-up. The experiments were carried out at ENEA Brasimone R.C. with the TRIEX-II facility on a mock-up characterized by a shell and tube configuration and using niobium as a membrane material. The experimental campaign was carried out with LiPb flowing at about 450 °C and 1.2 kg/s, while the hydrogen partial pressure was varied in the range 170–360 Pa. The characterization of the PAV performance was conducted by measuring the hydrogen partial pressure drop across the mock-up and the hydrogen permeated flux through a leak detector calibrated with an external hydrogen calibration cylinder. Moreover, the permeated flux was confirmed by a pressurization test performed measuring the pressure increase on the vacuum side of the PAV. The results constitute the first verification of the possibility to operate a PAV in flowing LiPb and to quantify its capabilities. [ABSTRACT FROM AUTHOR]

Tritium was extracted from tritium-contaminated aluminum samples by heating it in a high-temperature furnace at 200, 300, or 400 °C for 15 h. The extracted tritium was analyzed by using a liquid scintillation counter (LSC); the sample thicknesses were 0.4 and 2 mm. The differences in tritium extraction over time were also investigated by cutting aluminum stick samples into several pieces (1, 5, 10, and 15) with the same thickness, and subsequently heating them. The results revealed that there are most of the hydrated material based on tritium on the surface of aluminum. When the temperature was increased from 200 or 300 °C–400 °C, there are no large differences in the heating duration required for the radioactivity concentration to be lower than the MDA value. Additionally, at the same thickness, because the surface of aluminum is only contaminated to tritiated water, cutting the aluminum samples into several pieces (5, 10, and 15) did not have a substantial effect on the tritium extraction fraction at any of the applied heating temperatures (200, 300, or 400 °C). The proportion of each tritium-release materials (aluminum hydrate based on tritium) were investigated via diverse analyses (LSC, XRD, and SEM-EDS).

Tritium extraction from radioactively contaminated cement mortar samples was performed using heating and liquid scintillation counting methods. Tritiated water molecules (HTO) can be present in contaminated water along with water molecules (H2O). Water is one of the primary constituents of cement mortar dough. Therefore, if tritium is present in cement mortar, the buildings and structures using this cement mortar would be contaminated by tritium. The radioactivity level of the materials in the environment exposed to tritium contamination should be determined for their disposal in accordance with the criteria of low-level radioactive waste disposal facility. For our experiments, the cement mortar samples were heated at different temperature conditions using a high-temperature combustion furnace, and the extracted tritium was collected into a 0.1 M nitric acid solution, which was then mixed with a liquid scintillator to be analyzed in a liquid scintillation counter (LSC). The tritium extraction rate from the cement mortar sample was calculated to be 90.91% and 98.54% corresponding to 9 h of heating at temperatures of 200 °C and 400 °C, respectively. The tritium extraction rate was close to 100% at 400 °C, although the bulk of cement mortar sample was contaminated by tritium.

In a lead-lithium (PbLi) based breeding blanket, the bred tritium remains dissolved in the PbLi, which flows towards the Tritium Extraction Unit (TEU). In a TEU, tritium must be extracted at a fast enough rate that guarantees the plant's self-sufficiency and safety. Different TEU concepts exist, one of the most promising being Permeation Against Vacuum (PAV), based on the extraction of tritium from the PbLi through a highly permeable membrane. Even in the so-called low velocity blanket concepts, PbLi is expected to flow at relatively high total mass flow rates. This means that the high tritium extraction efficiencies that safe operation requires must be obtained partitioning the flow into several extraction channels, but this solution increases both cost and complexity. However, less partition channels may be required should turbulence be induced in the flow. Since turbulence increases the flow mixing, it should favor tritium extraction. Hence, in the range of velocities where turbulent phenomena start —i.e., the transition region—, extraction efficiency is expected to grow rapidly with velocity due to turbulence acting as a new transport mechanism. Thus, a turbulent flow in the TEU may achieve the high extraction efficiencies required with a more moderate partitioning scheme. In this work, a PbLi flow in a prototypical PAV channel was modeled using Direct Numerical Simulations (DNS). To trigger turbulence, two different methods were implemented, namely an instability-inducing oscillating boundary condition, and a physical turbulator consisting of a geometrical obstacle. Both methods proved successful as they resulted in greater extraction efficiencies than those seen in the analogous laminar regimes. In fact, up to a 15% increase in extraction efficiency or up to a 5-fold increase in total extraction rate were obtained. • The characteristic puffs of turbulence increase mass transfer through the boundary layer in the transition regime. • Turbulators and oscillating velocity boundary conditions are successful in triggering turbulence in the transition regime. • Extraction efficiency is improved when turbulence is triggered in the transition regime. • The transition regime stands as an interesting operational regime for a tritium extraction unit. [ABSTRACT FROM AUTHOR]

Effective tritium extraction from PbLi flows is a requirement for the functioning of any PbLi based breeding blanket concept. For a continuous plant operation, the removal of the tritium dissolved in the PbLi has to be performed in line and sufficiently fast. Otherwise, tritium inventories in the liquid metal, start-up inventories and buffer inventories would be excessive from the safety point of view. Moreover, a slow response of the tritium extraction systems could also compromise the tritium self-sufficiency of the plant. A promising solution to this problem is to use highly permeable membranes in contact with the PbLi flow to promote the extraction via permeation. This technique is usually known as Permeation Against Vacuum (PAV). As an alternative, tritium could be extracted directly by permeation through a fluid free surface (FS) in contact with vacuum. In both configurations, the dynamics of tritium transport is ruled by a combination of convection, diffusion and surface recombination. In this paper, the tritium extraction processes in the FS and PAV configurations are studied in detail. For the first time, general analytical expressions for the extraction efficiency are derived for both techniques in a Cartesian geometry. These expressions are general in the sense that they do not impose any kind of assumption concerning the permeation regime of the membrane or the fluid boundary layer. The derived expressions have been used to analyze numerically the response of both configurations in a close loop system, such as the one of DEMO. The presented methodology allows comparing the FS and PAV configurations, assessing in which conditions one will be behave better than other.

• Hydrogen Isotopes Extraction System (HIES) for extracting hydrogen from liquid Pb-Li is designed, fabricated and installed at IPR, India. • Liquid Pb-Li compatible structures packing (Sulzer BX type) is considered as gas-liquid contactor for extracting dissolved hydrogen from liquid Pb-Li. • Liquid Pb-Li flow distributor, gas distributor and assembly of structured packing in extractor column is presented in detail. • For effective dissolution of hydrogen gas in liquid Pb-Li, the gas sparger with bubbling nozzles tubes and its internal assembly with operation is discussed too. • Process design of extractor column including all process tanks is discussed. HIES 3D layout, piping flexibility and seismic analysis is also eloborated. Tritium breeding, recovery and safe storage are very important for achieving self-sustainability in a nuclear fusion reactor. Indeed, it is one of the essential task in terms of safety and fuel cycle aspects. Usually, tritium breeding is performed in the blanket module. In Indian blanket module, there are mainly two types of tritium breeder materials, one is Lithium based ceramic pebbles Li 2 TiO 3 (Lithium Titanate) as solid breeders and second is the lithium based alloy Pb-Li (Lead Lithium) as a liquid breeder. Tritium bred in solid breeders is extracted with purge helium gas, whereas tritium bred in liquid Pb-Li needs suitable gas-liquid contactor. It should be noted that, tritium recovery (i.e.,-extraction) from Pb-Li is very challenging due to low solubility of tritium in liquid Pb-Li and no such separation technology readily available commercially. Therefore, the necessity for developing a reliable and efficient tritium recovery technology (i.e.,- extractor) for liquid Pb-Li is of key interest. In present work, design and development of Hydrogen Isotopes Extraction System (HIES) for hydrogen recovery from liquid Pb-Li is discussed. As hydrogen and tritium are isotopes, hydrogen is used instead of tritium in present system due to several safety concerns with tritium. Recovery of hydrogen from liquid Pb-Li is carried out using structured packing as a Gas-Liquid Contactor (GLC). The design details of gas-liquid contactor (i.e.-extractor column) and various process tanks of HIES are discussed. In present design, configuration of vertically mounted extractor column makes overall HIES relatively tall (∼ 3 m height) and use of liquid Pb-Li as process fluid makes the system heavy weight (∼1500 kg) too. In addition to this, system needs to be operated above Pb-Li melting temperature (> 235 C) to keep Pb-Li in liquid form. Therefore, thermal analysis of piping as well as seismic analysis of whole system is carried out and discussed along with various operational stages of HIES. [ABSTRACT FROM AUTHOR]

• The turbulent two-phase model coupled with dilute mass transfer is adopted to numerically study the process of tritium extraction by gas–liquid contactor. • The numerical models include the single bubble, helium purge and the entire as-liquid contactor. • The total efficiency can achieve 95% when there are 13 units of gas–liquid contactor working together in series. As one key component of the fusion reactor, the blanket is in charge of breeding tritium, shielding neutrons and extracting thermal energy for electricity generation. An efficient tritium extraction unit (TEU) is a necessary guarantee for the tritium self-sufficiency in the fusion reactor. The gas–liquid contactor (GLC) is considered to be one of the most efficient devices for the liquid blanket that uses the lead–lithium (PbLi) as the functional materials. Understanding the characteristics of gas–liquid two-phase flow and mass transfer is beneficial for guiding the design and optimization of contactor. However, it is difficult to carry out the tritium experiments due to the scarcity in nature, the strong permeability, as well as the radioactivity. In this paper, the turbulent two-phase model coupled with dilute mass transfer is adopted to numerically study the process of tritium extraction from the liquid PbLi, and the numerical model is evolving step-by-step. Firstly, based on the models of single bubble and helium purge, the effects of different design parameters on the tritium extraction efficiency of the contactor are further investigated (i.e. mass flow rate of helium, initial tritium concentration and liquid PbLi filling height). The results are in good agreement with previous experimental results, and this validates the feasibility of the numerical model used in this work. Then, the entire model of GLC is numerically analyzed, which illustrates that the tritium extraction efficiency of one single unit is only 20.8%. However, the total efficiency can achieve 95% when there are 13 units working together in series. On this basis, the schematic diagram of tritium extraction system is proposed for the fusion blanket. [ABSTRACT FROM AUTHOR]

Tritium extraction from lithium-lead eutectic alloy (LiPb) is one of the biggest challenges to be solved for the exploitation of the WCLL (Water Cooled Lithium-Lead) as the Breeding Blanket (BB) of EU DEMO reactor. The Permeator Against Vacuum (PAV) is one of the most promising technologies to reach this goal, but it needs a bigger amount of experimental data to demonstrate its viability and to support the scale-up towards DEMO. For this reason, an experimental campaign was performed at ENEA Brasimone in the TRIEX-II facility, where PAV-ONE, a PAV mock-up based on niobium U-tubes, was characterized in flowing LiPb. This paper describes the results obtained at a temperature of 350 °C with a hydrogen partial pressure in LiPb in the range 110–230 Pa, conditions that are relevant for the WCLL BB. Furthermore, the results at 350 °C are compared with those previously obtained at 450 °C on the same mock-up and with the same flow rate, showing a reduction in the extracted hydrogen flux of about one order of magnitude. Finally, an analytical model was set up to reproduce the experimental results of the PAV mock-up. The model results had a closer fit to experimental results when using lower bound properties. This suggests that mass transport through the PbLi and molecular recombination from the PAV surface were the rate-limiting processes. [ABSTRACT FROM AUTHOR]

• Simulations and experimental uncertainties used to develop an experimental plan. • 6 Equilibrium pressures, 6 Pb-16Li temperatures, 2 nozzle geometries to be tested. • Theoretical extraction efficiencies >85% reported from simulations. 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. [ABSTRACT FROM AUTHOR]

A new facility, CLIPPER, is being constructed at CIEMAT to investigate tritium extraction from PbLi. It consists of a forced circulation loop with the main objective of validating the technique of permeation against vacuum. CLIPPER is designed as an isothermal loop operating up to 500 ℃. The PbLi is heated, melted and cleaned outside the loop, in a tank integrated in a dedicated glove box under argon atmosphere. The test section includes a prototype of permeator (TRITON) and its auxiliary systems, such as its associated vacuum system and instrumentation. TRITON consists of a rectangular multi-channel component with vanadium membranes. A novel gas injection system specifically designed for CLIPPER is used to introduce the gases that must be extracted through the permeator (hydrogen isotopes). This system, based on a multi-tube component with permeable niobium membranes, is able to solubilize the hydrogen in the PbLi up to the required concentration. The final design of CLIPPER is presented together with the integration of its main components, including a thermomechanical assessment of the loop. [ABSTRACT FROM AUTHOR]

Abstract In a fusion reactor based on liquid breeding blankets, tritium is generated due to the neutron irradiation of lithium based alloy, such as eutectic lead-lithium (PbLi). Then, tritium is extracted from the liquid metal by means of technologies that are presently under development such as the vacuum sieve tray, the permeation against vacuum or the gas liquid contactors. Nevertheless, for the experimental validation of these technologies at laboratory scale, hydrogen isotopes cannot be generated in situ in the liquid metal as in the breeding blanket. Hence, a system able to inject the gas in the flowing liquid at desired concentrations is required, avoiding the formation of bubbles as a consequence of the low solubility of hydrogen/deuterium in PbLi. The system should be capable of solubilizing the hydrogen to replicate as close as possible the conditions of a breeding blanket. A design of an injector based on a permeable membrane is here presented, being the driving force the gradient of concentrations existing between the two surfaces of the membrane. A hydrogen transport model for a tube-in-tube injector has been developed, showing that the injected hydrogen flux is proportional to the tube radius. However, the change on this parameter, that affects the velocity of the liquid and thus the mass transport coefficient, has opposite consequences on the rate of injection whose final impact relays on the properties of the employed membrane. An evaluation of the physical and geometrical aspects of a conceptual injector is depicted with the aim of optimizing the design to obtain adequate injection rate depending on the facility where the injector is installed. Finally, a conceptual design of a system to be implemented in an experimental PbLi loop for the validation of the permeation against vacuum technique for tritium extraction from PbLi is presented. The injector is based on a multi-tube component made of niobium able to inject hydrogen at the same rate as it is extracted under relevant conditions for a Dual Coolant Lithium Lead breeding blanket. [ABSTRACT FROM AUTHOR]

Abstract Based on the previous study in frontal displacement chromatography (FDC) packed with Pd-Al 2 O 3 , three groups of separation tests were carried out to verify the separation performance of the constituted FDC device for various compositions of feed gas and to validate the application probability of FDC in the Tritium Extraction System (TES) of ITER and China Fusion Engineering Test Reactor (CFETR). The separations were conducted by the FDC procedure with characteristics of the feed gas one-time flushing though the column and then reasonable separation performance had been obtained. The results indicate that the FDC process could be applied to deal with the desorbed gas mixtures from TES and/or further to extract and thereafter enrich the breeding tritium in ITER or CFETR, which would take the advantages of system compactness and efficiency over the present route of TES. Comparing to other related displacement chromatography procedures, the FDC process could be applied in tritium pre-enrichment for the mixtures of low tritium concentrations, which is highlighted by the outstanding merit of operation simplicity. Highlights • Various compositions of hydrogen isotope mixtures were separated by FDC. • Mixtures comparable to TES purge gas containing breeding tritium were separated. • The simulated separation results indicate FDC could be applied to simplify TES route in ITER. [ABSTRACT FROM AUTHOR]

Quantitative extraction of tritium from a sample matrix is critical to efficient measurement of the low-energy pure beta emitter. Oxidative pyrolysis using a tube furnace (Pyrolyser) has been adopted as an industry standard approach for the liberation of tritium (Warwick et al. in Anal Chim Acta 676:93-102, 2010) however pyrolysis of organic-rich materials can be problematic. Practically, the mass of organic rich sample combusted is typically limited to <1 g to minimise the possibility of incomplete combustion. This can have an impact on both the limit of detection that can be achieved and how representative the subsample is of the bulk material, particularly in the case of heterogeneous soft waste. Raddec International Ltd (Southampton, UK), in conjunction with GAU-Radioanalytical, has developed a new high-capacity oxygen combustion bomb (the Hyperbaric Oxidiser; HBO) to address this challenge. The system is capable of quantitatively combusting samples of 20-30 g under an excess of oxygen, facilitating rapid extraction of total tritium from a wide range sample types. [ABSTRACT FROM AUTHOR]

This paper describes the quantitative analysis for the design of a tritium extraction system that uses liquid PbLi droplets in vacuum (Vacuum Sieve Tray, VST), for application to the ITER helium-cooled lithium lead (HCLL) test blanket system (TBS). The parametric dependences of tritium extraction efficiency from the main geometrical features such as initial droplet velocity, nozzle head height, nozzle diameter, and flow rate are discussed. With nozzle diameters between 0.4 and 0.6 mm, extraction efficiency is estimated from 0.77 to 0.96 at the falling height of 0.5 m, with flow rate between 0.2 and 1.0 kg/s. The device has a height of 1.6 m, within the external dimensions of the HCLL Test Blanket Module (TBM), and no additional pumping power is required. The attained results are considered attractive not only for ITER, but also in view of the application of the VST concept as a candidate tritium extraction system for the European Union's demonstration fusion reactor (DEMO). The extraction efficiency of a single droplet column, which is the basis of the design analysis presented, has been validated experimentally with hydrogen. However, further experiments are required on an integrated system with size relevant to the proposed HCLL-TBS design to validate system-level effects, particularly regarding the desorption process in an array of multiple droplets. [ABSTRACT FROM AUTHOR]

Europe is committed in developing the design of the two Test Blanket Systems (TBS) based on HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed) breeding blanket (BB) concepts. The complexity of the TBS design comes not only from the innovative fabrication technologies and materials adopted for Test Blanket Modules (TBM) but also from the requirements and functions that the TBM ancillary systems have to satisfy and implement. Indeed, the main TBM ancillary systems, namely the Helium Cooling System, the Coolant Purification System and Tritium Extraction System, all belonging to the Safety Important Class (SIC), have to implement fundamental functions, like the transport of the surface and volumetric heat from the TBM to the heat sink, the extraction and processing of the tritium generated in the TBM, the confinement of radioactive inventory, the support to the investment protection and safety functions. On top of the full compliance with the ITER safety principles, the design of the TBM systems is focused on providing high operational reliability and availability not to jeopardize ITER program and, at the same time, also a good operational flexibility to make possible the achievement of the main TBM scientific objectives. This paper gives an overview of the design status of the HCLL and HCPB-TBM (ancillary) systems, updated to the conclusion of the conceptual design phase (CDR). The most relevant technologies, the still open points, the main issues related to the integration in ITER and last relevant results from the on-going R&D activities carried out in Europe are presented and discussed. In the last part, different considerations are proposed about the impact of the design and operation of the main HCLL and HCPB-TBM ancillary systems technologies on the design of a DEMO BB. [ABSTRACT FROM AUTHOR]

The design of a DEMO reactor requires the design of a blanket system suitable of reliable T production and heat extraction for electricity production. In the frame of the EUROfusion Consortium activities, the Breeding Blanket Project has been constituted in 2014 with the goal to develop concepts of Breeding Blankets for the EU PPPT DEMO; this includes an integrated design and R&D programme with the goal to select after 2020 concepts on fusion plants for the engineering phase. The design activities are presently focalized around a pool of solid and liquid breeder blanket with helium, water and PbLi cooling. Development of tritium extraction and control technology, as well manufacturing and development of solid and PbLi breeders are part of the programme. [ABSTRACT FROM AUTHOR]

Hydrogen isotopes solubility in eutectic lithium–lead alloys is really important for the design of breeding blanket components that use this breeding material. The determination of the magnitude and kinetics of the tritium flux from the blanket to the helium cooling loop, along with the design of future tritium extraction systems of the breeding alloy or the He coolant purification system, will be defined on basic transport parameters such as solubility. The unacceptable scattering of Sieverts’ constant values in the historical measurements given by different experimental techniques, suggests that this is a very important and unresolved issue. In this work, it has been experimentally evaluated, using absorption and desorption techniques. The different measurement campaigns have been carried out in the temperature range from 523 to 922 K and in the pressure range from 1 to 10 5 Pa. This paper describes the work carried out in the preparation of the facility, the theoretical model developed to process the different results obtained by means of absorption and desorption runs. Final results obtained during several campaigns of measurements are provided. The obtained values of hydrogen solubility through the different campaigns show a similar value for the Sieverts’ constant, and therefore, a very little value for the activation energy in the solution process. Results are compared and discussed. The proposed correlation for hydrogen Sieverts’ constant in Pb–Li from these tests is K S [mol m −3 Pa −1/2 ] = 8.64 × 10 −3 exp(–0.9/ RT ), R in (kJ K −1 mol −1 ). [ABSTRACT FROM AUTHOR]

The model of the desorption of hydrogen isotopes from lead lithium alloy in a packed column is derived from the first principles using the plug flow in the liquid phase either the plug flow or ideal mixing in the gas phases. Sievert's law of non-linear equilibrium is followed. The volumetric mass transfer coefficient k L a and its dependence on the liquid metal flow rate are evaluated on the basis of the Melodie loop experiments. The presented model is used for evaluation of the minimum flow rate of the purge gas for which the concentration of the isotope in the gas leaving the column is at its highest, while the driving force of the interfacial transport of the isotope is still not reduced and the tritium desorption efficiency is therefore retained. The potential effect of the axial dispersion in the gas and liquid phase is evaluated. Highlighted are the issues of the optimum packing geometric surface area, above which the efficiency starts to decrease, and of the role of the surface tension and the contact angle with regard to the wettability of the packing. On the basis of the findings related to these factors, the Mellapak 500 Y and Mellapak packings with flat surfaces are recommended for the tests aiming to intensify the tritium desorption efficiency in the packed columns. The models were used for the engineering sizing of the packed columns in two breeding blanket concepts for the DEMO plant – utilizing DCLL (dual coolant lead lithium) and HCLL (helium cooled lithium lead). [ABSTRACT FROM AUTHOR]

Abstract: The HCPB (Helium Cooled Pebble Bed) Test Blanket Module (TBM), developed in EU to be tested in ITER, adopts a ceramic containing lithium as breeder material, beryllium as neutron multiplier and helium at 80bar as primary coolant. In HCPB-TBM the main function of Tritium Extraction System (TES) is to extract tritium from the breeder by gas purging, to remove it from the purge gas and to route it to the ITER Tritium Plant for the final tritium processing. In this paper, starting from a revision of the so far reference process considered for HCPB-TES and considering a new modeling activity aimed to evaluate tritium concentration in purge gas, an updated conceptual design of TES is reported. [Copyright &y& Elsevier]

Abstract: Tritium extraction system is essential for CN HCSB TBM for safety and technical reasons. Based on the assessments of system functions, integration issues and safety considerations, two main modifications of the system from previous design (Feng et al., 2007 ; Chen et al., 2008 ) are adopted: [a)] the TES has been split to 2 parts with one in port cell and another in tritium building. Q2O in the purge gas is reduced to Q2 in a hot metal bed located in port cell; Q2 is separated from the stream by a pair of cryogenic molecular sieve beds and a Pd/Ag diffuser located in tritium building. [b)] isotope separation process has been excluded. TES components sizes are estimated and space allocations are estimated. Required services and where and when they are needed are preliminary defined. Fluids delivered towards ITER tritium system are analyzed. [ABSTRACT FROM AUTHOR]

Abstract: The conceptual design of the ancillary systems of the European Test Blanket Modules (TBMs), namely the PbLi circuit, the Helium Cooling Systems (HCSs), the Coolant Purification Systems and Tritium Extraction Systems (TESs), is proceeding as per time schedule adopted by the European Domestic Agency for ITER (Fusion for Energy (F4E)). A general description of these systems, based on the present baseline, is given in this paper. Of basic importance for the conceptual design of the ancillary systems is the definition of the main interfaces with ITER because of the potential impact not only on their design itself but also on their operation and maintenance strategy. The main interfaces and integration issues in ITER involving the ancillary systems are here described and discussed. Moreover, the most critical future activities and the main milestones, identified to be compliant with the objective to deliver to ITER the European TBS ancillary systems by 2019, are also indicated and discussed in this paper. [ABSTRACT FROM AUTHOR]

Abstract: Effective tritium breeding achievable in Test Blanket Module (TBM) is a major issue for sustainable fusion energy program. Equally important is tritium extraction to recover and recycle tritium back to fusion reactor. Tritium extraction from lead lithium is much more complicated than from purge gas due to low tritium extraction efficiency in transfer step to gas phase and the limitations imposed on space and lead lithium inventory in port cell. Earlier investigations do suggest the preference of packed columns over bubble columns. Theoretical models based on axial dispersion plug flow in liquid and gas proposed for bubble columns and packed columns are reinvestigated for different boundary conditions. This paper highlights the critical issues of experimental design based on tritium extraction efficiency and its impact on recovery loop. Steady state closed loop for absorption and stripping of hydrogen isotopes using inert gas is designed along with the associated auxiliaries. [ABSTRACT FROM AUTHOR]

Abstract: One of the most challenging issues for testing the different Test Blanket Modules (TBM) concepts in ITER is the demonstration of the ability to correctly and efficiently manage the bred tritium. For all TBM concepts the main auxiliary systems involved in this task are the Tritium Extraction System (TES), whose main purpose is to extract tritium from the breeder, making it available to the Tokamak Exhaust Processing system in the ITER Tritium Plant, and Coolant Purification System (CPS) which extracts the permeated tritium from He-coolant and keeps controlled the chemistry of the primary cooling circuit. Both these systems have to be compatible with: [-] the foreseen requirements from the TBM experimental campaign; [-] the ITER space requirements; [-] the requirements in terms of interface with the ITER Tritium Plant. In this paper, the composition and flow-rate of the gas stream to be processed by the tritium extraction system and coolant purification system in the Helium Cooled Pebble Bed Test Blanket Module is given for the high duty DT phase of ITER. Moreover, in the light of the operative conditions planned for the experimental campaign on the HCPB-TBM, a preliminary design of the TES and CPS is presented and discussed, with a first estimation of the required space for their installation. [Copyright &y& Elsevier]

Main results of in-pile reactor testing of the first model of the tritium breeding zones (TBZ) for fusion reactor blankets containing both lithium orthosilicate and beryllium spheroid particles are presented. Experimental results of tritium inflow - in the form of HT and HTO - into neon purge-gas with or without 1% hydrogen are discussed. [ABSTRACT FROM AUTHOR]

The eutectic Pb-Li alloy is used as liquid breeding blanket material in the concept design of future fusion nuclear reactors where the effective extraction of generated tritium is an important unit operation of the fusion fuel cycle. In this context, the adoption of porous stainless steel tubes working as Membrane Gas-Liquid Contactors (MGLCs) has been proposed for recovering hydrogen isotopes from liquid Pb-Li alloys. Taking into account the results achieved in a previous experimental campaign carried out at 370 °C, the hydrogen mass transfer coefficient has been evaluated and compared with data available in literature. The present study reveals that hydrogen extraction consists of diverse steps, although it is mainly controlled by recombination of hydrogen atoms and their desorption from the gas-liquid interface. The overall permeation exhibits mass transfer resistance values in the range 10−7 – 10-6 m2 s Pa mol-1, while hydrogen transport through the MGLC pores takes place with mass transfer resistances significantly smaller (around 10-4 m2 s Pa mol-1). These results make the use of MGLCs very promising for an effective extraction of tritium from liquid breeders in fusion reactors. [ABSTRACT FROM AUTHOR]

• Design, optimization and analysis of a liquid nitrogen cryostat system is performed. • The design is carried out based on ASME Section VIII Div-1 and 2. • The thermal fluid structure interaction analysis of the system is performed using ANSYS. • The cryostat is fabricated and successfully tested at operating and design conditions. Efficient design of Tritium Extraction System (TES) for the fuel cycle of any fusion reactor is very important to maintain the tritium breeding ratio and hence sustain the fusion reaction. Hydrogen Isotopes Removal System (HIRS) for Indian Tritium Breeder Blanket removes Q 2 (Q = H, D or T) and impurities using Cryogenic Molecular Sieve Bed (CMSB) adsorber at 77 K. The CMSB is maintained at liquid nitrogen temperature using a double walled cryostat made up of SS304. The paper describes the design and thermal Fluid Structure Interaction (FSI) analysis of cryostat assembly for CMSB of HIRS. The coupled analysis performed in this work involves solving for the fluid domain and transferring the results to ANSYS Thermal-Static structural set up to determine the stresses and displacement due to combined effects in the system. The mechanical design of the cryostat components is analytically performed using ASME codes. The velocity, pressure drop and time taken to cool the CMSB are determined by solving the fluid and energy equations in ANSYS Fluent analysis system. The solutions are imported into ANSYS Thermal-Static structural analysis system and the thermal-structural stresses and deformations are determined considering the temperature, pressure and acceleration loads. The space between inner and outer vessel is maintained at vacuum, which might lead to buckling. So, the critical buckling load multiplier factor is determined. These results are used in fabricating the complete cryostat system for CMSB of HIRS. [ABSTRACT FROM AUTHOR]