Tritium Extraction System Research Articles

Design and Integration of the EU-DEMO Water-Cooled Lead Lithium Breeding Blanket

The water-cooled lead lithium breeding blanket (WCLL BB) is one of two BB candidate concepts to be chosen as the driver blanket of the EU-DEMO fusion reactor. Research activities carried out in the past decade, under the umbrella of the EUROfusion consortium, have allowed a quite advanced reactor architecture to be achieved. Moreover, significant efforts have been made in order to develop the WCLL BB pre-conceptual design following a holistic approach, identifying interfaces between components and systems while respecting a system engineering approach. This paper reports a description of the current WCLL BB architecture, focusing on the latest modifications in the BB reference layout aimed at evolving the design from its pre-conceptual version into a robust conceptual layout. In particular, the main rationale behind design choices and the BB’s overall performances are highlighted. The present paper also gives an overview of the integration between the BB and the different in-vessel systems interacting with it. In particular, interfaces with the tritium extraction and removal (TER) system and the primary heat transfer system (PHTS) are described. Attention is also paid to auxiliary systems devoted to heat the plasma, such as electron cyclotron heating (ECH). Indeed, the integration of this system in the BB will strongly impact the segment design since it envisages the introduction of significant cut-outs in the BB layout. A preliminary CAD model of the central outboard blanket (COB) segment housing the ECH cut-out has been set up and is reported in this paper. The chosen modeling strategy, adopted loads and boundary conditions, as well as obtained results, are reported in the paper and critically discussed.

Low-pressure adsorption of hydrogen isotopologues on LTA4A zeolites - A grand canonical Monte Carlo simulation study

Effective removal of the trace quantities of hydrogen and its isotopologues is crucial for the proper functioning of the tritium extraction system (TES) in the fuel cycle system of the fusion reactor. Zeolite molecular sieves of Linde Type A (LTA) framework are considered potential adsorbent materials for the cryogenic molecular sieve bed (CMSB) of the TES. However, the adsorption isotherm data of H2, HD, D2, HT, DT, and T2 on LTA 4A zeolites at low pressures (1–1000 Pa) and 77.4 K temperature is scarce in the literature. Grand canonical Monte Carlo (GCMC) simulations were performed to study the hydrogen isotopologues adsorption in the pressure and temperature ranges of operation of the CMSB of TES. The adsorption isotherms show that the equilibrium loading capacity is larger for T2, followed by D2 and H2 at 77.4 K. In the case of heteronuclear isotopologues, the loading capacity is larger for DT, followed by HT and HD under the same conditions of interest. The increase in the equilibrium adsorption capacity with isotopic mass could be attributed to the difference in the zero-point energies of the isotopologues. The effect of Si/Al ratio in the LTA zeolites on the cryosorption of hydrogen isotopes is investigated. The results show that, as the Si/Al ratio decreased, the equilibrium loading of the hydrogen molecules increased. The results obtained from this work are in good agreement with our experimental results and published data in the literature.

Compatibility of niobium, vanadium and P22 steel in high temperature flowing LiPb

The LiPb loops of the WCLL BB (Water Cooled Lithium-Lead Breeding Blanket) have the primary aim of routing tritium-rich LiPb from the Breeding Blanket to the Tritium Extraction System of DEMO Reactor. Besides, the loops maintain LiPb in the desired conditions and purify it from corrosion and activated products and from helium generated by the neutron transmutation of lithium. For some of these tasks the loops need special functional materials. Among them, niobium and vanadium are widely considered for the tritium extraction technologies for their high permeability to tritium. In addition, the loops will be made of structural materials capable of withstanding the operative conditions and the aggressive environment. The current reference material for the piping is the ASTM A335 Grade P22. This paper investigates the compatibility at 500 °C of niobium, vanadium and P22 with LiPb flowing at 0.5 m/s. This velocity is the maximum foreseen in the LiPb loops, while the temperature was chosen following a conservative approach, as the maximum expected temperature is 450 °C in the BB and 330 °C in the LiPb loops. The specimens of the three materials were exposed for 1,000, 2,000 and 4,000 h in IELLLO LiPb facility at ENEA Brasimone Research centre. After the exposure, the specimens were examined by SEM-EDX, both in cross-section and surface view, XRD analyses were performed in order to investigate potential corrosion compounds formed. Weight loss analyses were also performed on niobium specimens, after chemical cleaning from adherent LiPb. The analyses highlighted the impact of corrosion products and their interaction with refractory metals (Ni and Fe) and P22 steel (Cr, after redeposition).

Preliminary accident analyses of ex-vessel LOCA and TES pipe break for CFETR HCCB blanket system

Safety analysis is one of the most important parts in the design of nuclear power reactor. It is a necessary step for China Fusion Engineering Test Reactor (CFETR) to make the research in safety accidents analysis of main coolant systems and tritium systems. The consequences of nuclear accidents can be used to prove the safety of plant and optimize the systems design. Now, the main design parameters of CFERR have been determined, and the conceptual design of the main systems have been done. This paper presents two design basis accidents (ex-vessel loss of coolant accident (LOCA) and TES pipe break) analysis results in primary helium cooling system (HCS), and tritium extraction system (TES). According to the accident analysis results, the radioactive release assessments are given. The analysis results can provide reference for the following parameter optimization design of CFETR systems.

Tritium transport code development based on COMSOL and MATLAB and its application to the WCCB blanket

Water cooled ceramic breeder (WCCB) blanket was proposed as one of the candidate blanket schemes of the Chinese Fusion Engineering Test Reactor (CFETR). In this paper, a code simulating tritium transport in WCCB blanket and its external loops has been developed, where zero and two dimensional models are integrated to get more reliable results. The purge gas flow in the breeder zone, heat transfer in the blanket and tritium isotopes mass transport are considered. The isotope exchange in the purge gas and coolant loop are also taken into account in the model. The two dimensional model of the WCCB blanket and zero dimensional model of external loops are integrated to calculate tritium related quantities. The amount of tritium inventory in different components, tritium losses, tritium extracted by the tritium extraction system (TES) and purified by the coolant purification system (CPS) are obtained. The sensitivity analysis about key parameters has also been conducted.

Exploring the use of electrochemical hydrogen pump in tritium extraction system and coolant purification system

The advanced tritium extraction system (TES) and coolant purification system (CPS) is designed to directly separate hydrogen isotopes containing tritium from purge gas and coolant gas at high temperatures (773–873 K). In order to explore whether the cogeneration can be fully used for preheating at this temperature to make the hydrogen pump (HP) work normally, the electrochemical HP based on chemically-stable BaZr0.8Y0.16Ni0.04O3-δ (BZYN) protonic electrolyte and SrFe0.8Mo0.2O3-δ (SFM) symmetrical electrodes is proposed. Hydrogen was almost completely extracted from a mixed gas with 0.1% hydrogen when the voltage of only 1.50 V was applied on the protonic HP at 873 K. And hydrogen was extracted from a mixed gas with 10 vppm hydrogen when the applied voltage was about 15.16 V at 873 K. The results have suggested that the electrochemical hydrogen pump based on BZYN is promising for advanced TES and CPS in fusion reactor.

The European Dual Coolant Lithium Lead breeding blanket for DEMO: status and perspectives

During the last years CIEMAT has been leading the activities in the European Program to develop an integral breeding blanket with advanced performances to work in a realistic DEMO scenario. This blanket is the dual coolant lithium lead (DCLL) working at a limited temperature in order to allow the use of conventional materials and technologies. The design of this blanket was finished, including the definition of the tritium extraction system and tritium simulations. Then, determined by the selection of other two concepts as driver blankets for DEMO, the focus was put on developing a novel BB still based on the DCLL concept but working at higher temperatures, thus increasing the plant net efficiency. In this work, a summary of the status of the DCLL is presented, together with some ideas for developing an advanced high temperature DCLL in the near future.

Study and characterization of potential adsorbent materials for the design of the hydrogen isotopes extraction and analysis system

One of the most challenging tasks in the design of the inner fuel cycle system lies in the effective design of Tritium Extraction System (TES), which involves proper extraction and purification of tritium in a fusion reactor. A prototype Hydrogen Isotopes Recovery System (HIRS) is being developed to validate the concepts for tritium extraction by adsorption mass transfer mechanism. The two main systems of HIRS are Atmospheric Molecular Sieve Bed (AMSB) adsorber and Cryogenic Molecular Sieve Bed (CMSB) adsorber. AMSB removes ppm levels of water vapour while CMSB removes ppm levels of hydrogen isotopes, oxygen and nitrogen gas from Helium purge gas. Selection of appropriate adsorbents for the HIRS is important for its efficient functioning. Adsorbents, namely, Zeolites 3A, 4A, 5A, 13X and Activated Carbon have been studied in detail at cryogenic temperatures to understand their surface characteristics and adsorption potential. In this work, we have generated the adsorption isotherms for hydrogen isotopes on potential adsorbents using a volumetric adsorption apparatus for a wide range of pressure from 1 Pa to 105 Pa. The BET and DFT models are applied for the determination of the textural properties of the adsorbents and the Langmuir-Freundlich composite model is used to fit the isotherm data. The parameters of the model were determined using nonlinear regression analysis. The enthalpy of adsorption is determined and is in the range of 8–11.5 kJ/mol for the adsorbate loadings corresponding to the partial pressure of hydrogen isotopes of ∼ 50–500 Pa in the TES. At low pressures of hydrogen isotopes, the free energy plot followed Henry’s law. The isotherms and enthalpy of adsorption clearly indicate reversible physisorption and monolayer formation of hydrogen isotopes on the potential adsorbents. Amongst hydrogen and deuterium, hydrogen is more strongly adsorbed on Zeolites 4A, whereas deuterium is more strongly adsorbed on Zeolites 13X. In case of activated carbon, no isotopic selectivity was observed. This study facilitates the selection of potential adsorbents for the CMSB and quantification of the parameters associated with the adsorbents.

Engineering design of a Permeator Against Vacuum mock-up with niobium membrane

Permeator Against Vacuum (PAV) is one of the technologies proposed for the Tritium Extraction and Removal System (TERS) of the Water-Cooled Lithium Lead Breeding Blanket (WCLL BB). The paper presents the activity aimed at the engineering design of a PAV mock-up with a niobium membrane, in order to later assemble and qualify it. Experience gained in the engineering design of the mock-up, the heating system, the instrumentation, and the vacuum line is illustrated. This experience will be useful for the preliminary design, the manufacturing and the operation of the PAV with niobium membrane for DEMO. Niobium was selected as membrane material of this mock-up because of its high permeability and for its lower cost compared to vanadium, the other candidate material for membranes. Besides, niobium has a lower tendency to oxidation than vanadium. Oxidation would reduce the hydrogen isotopes permeation flux. In this paper, the solution adopted to manufacture the PAV mock-up, a complex component with niobium and P22 parts, is illustrated. The Nb/P22 welding issues are also presented, in particular related to the compatibility of the welded joints with LiPb. In the chosen design, the LiPb flows with two passages in 16 (8 + 8) niobium “U” shaped pipes installed in a vacuum chamber and welded to a P22 plate. The U-pipes configuration was selected to minimize the welding area, the volume of the component and the membrane thickness while trying to preserve the highest possible extraction efficiency.

Membrane reactor performances in H2 recovery under conditions relevant for the HCPB tritium extraction system

The Tritium produced in the HCPB blanket is released in both elemental and oxidized form (the HT/HTO release rate is about 97/3). The reference process for removing the produced tritium relies on the use of Reactive Molecular Sieve Beds (RMSBs) and Cryogenic Molecular Sieve Beds (CMSBs). The open issues of such process are related to the regeneration procedure to be adopted in the RMSBs and on the suitable technique required for treating the gas stream leaving the RMSBs during the regeneration phase, which contains the released Q2O mixed with the He and H2 stream used for flushing the sieving beds. Membrane reactors represent a valid option due to their ability in performing Water Gas Shift (WGS) reaction and hydrogen separation directly in one device.In this study, the performances of a membrane reactor (a Pd/Ag membrane filled with 16 g of a commercial catalyst named ShiftMax® 120 HCF) in H2 recovery via Water Gas Shift reaction have been explored by performing both experimental tests and simulation analysis. Particularly two sets of experiments, differing solely in the H2O amount fed in the membrane reactor (5 and 10 g h-1), have been performed. In both, the H2O/CO feeding ratio was equal to 1 and four different conditions were investigated: i) no addition of H2 and He in the WGS feeding stream; ii) addition of H2 to the WGS feeding stream at several H2/H2O feed molar ratio; iii) addition of H2 and He to the WGS feeding stream at several H2/H2O and He/H2O feed molar ratio, iv) addition of He to the WGS feeding stream at several He/H2O feed molar ratio. The experimental results have been used to validate a novel simulation tool and in most of the cases a good agreement have been observed among the experimental and the simulation results.

Liquid metal MHD flow influence on heat transfer phenomena in fusion reactor blankets

DEMO reactor is projected as a technological demonstrator and a component test reactor for several key systems such as the breeding blanket. The dual coolant lithium lead (DCLL) is one promising breeding blanket candidate that performs the functions of shielding, part of the cooling and breeding with the same liquid metal fluid. PbLi is also in charge of transporting the bred tritium to the Tritium Extraction System. Structural cooling of the blanket walls is performed by pressurized helium flow which also transfers heat to the balance of plant. Keeping a proper distribution of thermal loads between the two coolants is a key aspect to reach a plant high efficiency. The thermal power transferred from the liquid metal to the helium channels depends on the flow regime. While liquid metal flows under the magnetic field, the magnetohydrodynamic (MHD) phenomena will occur, influencing the flow regime and the velocity profile. The heat is volumetrically deposited in the liquid metal coolant due to the interaction between the highly energetic neutrons originated in the plasma and the coolant. The heat deposition profile is roughly exponential causing temperature differences across the same channel cross section. Temperature gradients induce buoyant effects altering the expected MHD profile and affecting the heat transfer rate from the liquid metal to the surrounding walls cooled with helium. It is of utmost interest to identify the effects of the main flow parameters on the heat transfer coefficient under such complex conditions. The work studies computationally a domain consisting of a liquid region and a surrounding solid region electrically and thermally coupled using latest EU DCLL design characteristics. A fully developed flow is assumed where buoyancy is modelled under the Boussinesq approximation. Joule effect has been considered negligible compared to the neutron deposition. The influence of the buoyant MHD phenomena to the heat transfer coefficient is analysed parametrically with a CFD solver implemented in OpenFOAM®. The parameters investigated here are the Reynolds, Hartmann and Grashof numbers, the Grashof ratio, and the wall conductance ratio (cw).

Hydrogen separation from hydrogen/helium mixtures under different pressures by using U-shaped ZrCo getter beds

Helium is used to clean the torus and extract tritium produced by bleeding blanket in fusion facility. So, large amounts of hydrogen/helium mixtures would be produced during the operation of fusion facility and need to be separated for reuse. Cryogenic absorption bed is always used to realize this purpose along with Pd membrane for deep purification, which is high energy consuming and the procedure is complicated. ZrCo can absorb hydrogen selectively, ensuring its possibility for separating hydrogen from hydrogen/helium mixtures. So, ZrCo getter bed is proposed as a candidate process for the tritium extraction systems of TBM. However, the experimental data about this process is quite limited. In this work, experiments of gas mixtures with different compositions (30 %, 50 % and 80 % He) passing through a U-shaped ZrCo getter bed under different operation pressures (125, 150, 175 and 200 kPa) have been carried out. In general, the gas flux will increase when the operation pressure becomes higher for a given gas mixtures. When the pressure is fixed at a constant value, the flux of the higher hydrogen concentrated gas mixture is larger. Breakthrough always appears when hydrogen absorption volume reaches 30 %∼40 % of maximum equilibrium capacity of ZrCo bed in the static absorption. The obtained results in this paper can provide useful information for the practical design of ZrCo getter bed for H2/He separation in fusion facility or H2/inert gas separation in other situation.

Conceptual design of a PAV-based tritium extractor for the WCLL breeding blanket of the EU DEMO: Effects of surface-limited vs. diffusion-limited modeling

In order to close the fuel cycle, the EU DEMO will be equipped with a breeding blanket (BB) for the on-site tritium production. In the water cooled lithium lead (WCLL) BB concept, the tritium will be generated in the liquid PbLi from which it must then be extracted.The conceptual design of the tritium extraction and removal system for this BB concept is addressed here. It relies on the permeator against vacuum (PAV) technology, where the tritium dissolved in the PbLi is forced to cross a metal (Nb) membrane by pumping the vacuum on the secondary side. The proposed PAV geometrical configuration is based on the shell-and-tube concept, where the walls of the pipes in which the PbLi flows are the permeator membrane. The dimensioning of one permeator vessel for one of the PbLi loops is carried out here, adopting two different models for the tritium permeation: diffusion-limited (for the case of a non-contaminated membrane) and surface-limited (suitable for a contaminated Nb surface).The results show that in the case of a non-contaminated membrane it is possible to reach the target tritium extraction efficiency also at the PbLi nominal operating temperature, while in the case of a contaminated membrane the PbLi should be heated to enhance the permeation.

Conceptual design of the main Ancillary Systems of the ITER Water Cooled Lithium Lead Test Blanket System

The Water Cooled Lithium Lead Test Blanket System (WCLL TBS) is one of the EU Test Blanket Systems candidate for being installed and operated in ITER. In view of its Conceptual Design Review by F4E and ITER Organization (IO), planned for mid-September 2020, several technical activities have been performed in the areas of WCLL TBS Ancillary Systems design.In this article the outcomes of the conceptual design phase of the four main Ancillary Systems of WCLL TBS, namely the Water Cooling System (WCS), the Coolant Purification System (CPS), the PbLi loop and the Tritium Extraction System (TES), are reported and critically discussed.In particular, for each Ancillary System hereafter are reported: i) a short design description, including the conceptual design of their main components together with their operative conditions under the so-called Normal Operational State (NOS), ii) the ESP-ESPN classification for their main components, and iii) their arrangement and integration in the assigned ITER areas (PC#16, Vertical Shaft, TCWS Vault, Galleries and Tritium Process Room).

Experimental measurements of gas pressure drop of packed pebble beds

The nearly spherical shaped ceramic pebble beds in the breeder blanket of the future fusion reactor are purged by low-pressure gas to channelize the produced tritium fuel into the tritium extraction system. The required pumping power for the flowing gas in pebble beds can be estimated using the pressure drop across the pebble beds. The aim of this work is to measure the gas pressure drop experimentally across packed pebble beds as a function of pebble sizes, pebble shapes, pebble materials, and gas velocity. The pebble beds are packed in a cylindrical-shaped stainless steel container with an inner diameter of 24 mm and a length of 130 mm. The various experiments have been performed on stainless steel spheres (Diameter: 1 mm, 2 mm, 3 mm, and 4 mm), alumina pebbles (Mean diameter: 1 mm and 1.5 mm), and lithium meta-titanate pebbles (Mean diameter: 1 mm and 1.3 mm). The gas flow has been controlled and measured using a digital mass flow controller. The static differential pressure across the pebble beds has been monitored by a differential pressure transducer. The pressure drop significantly increases with a decrease in the diameter of pebbles/spheres and an increase in the packing fraction of the bed. The material type does not affect the results which are too obvious for the fixed pebble bed which is considered in these experiments. The obtained experimental results of gas pressure drop have been compared and agreed well with the prediction of the Ergun’s correlation.

A new integrated SSCs safety classification method and a case study on CN HCCB TES

Systems, Structures and Components (SSCs) have different safety significance for reactor operation, and it is necessary to classify SSCs into different safety classes for design, manufacture and maintenance of a reactor. In order to further identify the safety classes of SSCs in reactors, a new integrated safety classification method was proposed in this study, based on the 50-SG-D1 safety classification method, risk-informed safety classification method and integrating relative importance and functions importance classification. To demonstrate the advantages of the proposed method, the China Helium Cooled Ceramic Breeder Tritium Extraction System (CN HCCB TES) in the International Thermonuclear Experimental Reactor (ITER) was analyzed. New safety classes with more comprehensive information were obtained by applying the proposed method, indicating that the new safety classes of the items obtained by the proposed safety classification method could be of more significances than those derived from the present safety classification method.

Developments on the tritium extraction and recovery system for HCPB

Abstract The tritium release from the ceramic materials of the DEMO Breading Blanket (BB) is a complex mechanism due mainly to the chemistry of the interface between the breeder material and the purge gas. The tritium atomic form produced inside the ceramics structure is converted into tritiated water at the surface of the breeder material due to interaction with residual moisture / water layer on the ceramics and free oxygen atoms after Li burn-up. Therefore, in contrast to the liquid BB where tritium is produced as Q2 solely, the Tritium Extraction and Recovery (TER) from any solid BB must handle both the molecular tritium and the tritiated water. Most of the processes considered for T extraction from the purge gas foresee to handle Q2O and Q2 separately and consecutively, making necessary a two-stage process for full tritium removal.

Progress of the conceptual design of the European DEMO breeding blanket, tritium extraction and coolant purification systems

Abstract In the frame of the EUROfusion consortium activities the Helium Cooled Pebble Bed (HCPB) and the Water Cooled Lithium Lead (WCLL) concepts are being developed as possible candidates to become driver Breeding Blanket (BB) for the EU DEMO, which aims at the tritium self-sufficiency and net electricity production. The two BB design options encompass water or helium as coolants and solid ceramic with beryllium/beryllides or PbLi as tritium breeder and neutron multipliers. The BB segments have evolved towards a more stable conceptual design taking into account multiple feasibility aspects and requirements imposed by interfacing systems. Possible solutions to improve shielding capabilities of Helium cooled BB are investigated and the impact of water coolant activation is assessed by studying the spatial distribution of 16N and 17N isotopes dose rates, in particular in proximity of isolation valves. The reference and back-up technologies for the Tritium Extraction and Removal (TER) from the helium purge gas and the PbLi are developed addressing key feasibility aspects and implications on the tokamak layout and with considerable R&D efforts. As the BB internals offer an ideal environment (high temperatures, thin structural material) to promote the tritium permeation, studies are devoted to establish a tritium balance in the different systems during operation, with special care to the permeation rate and inventory in the coolant. Those are the key parameters for the feasibility assessment and technology selection for the Coolant Purification Systems (CPS).

Characterization of Pb-15.7Li Hydrogen Isotopes Permeation Sensors and Upgrade of Hyper- Quarch Experimental Device

The water-cooled lithium-lead (WCLL) test blanket module (TBM) of ITER and the respective breeding blanket (BB) of DEMO need to be efficiently characterized from the point of view of hydrogen isotopes inventory. An in situ detection is required to guarantee tritium self-sufficiency and to control the radiological hazards toward the external environment. In this context, a reliable hydrogen isotopes permeation sensor is essential both for tritium management at the several steps of the reactor fuel cycle and for monitoring the tritium processing systems, e.g., the tritium extraction system (TES). Hence, the hydrogen isotopes permeation sensors were characterized by protium experiments in a 1:1 scale experimental facility named TRIEX-II, located in ENEA C. R. Brasimone, Italy. The results obtained were used for preliminary validation of the novel hydrogen isotopes transport (HIT) code. Moreover, in order to test the sensor in different operative conditions, a new experimental device named Hyper-Quarch II (Hydrogen Permeation Quartz Chamber) is under development and will be installed in ENEA C. R. Brasimone, on the basis of the experience gained in the past experimental campaigns. This device, characterized by an upgraded quartz chamber and new instrumentation equipment, will be used to test and qualify an advanced hydrogen permeation sensor at a laboratory-scale in both dynamic and equilibrium modes: protium or deuterium will be used instead of tritium to simulate the operative conditions of ITER and DEMO reactors.

Sensitivity Analysis and Dimensioning of Reactor-Scale Pd/Ag Permeators for the Tritium Extraction and Removal System of the EU-HCPB Blanket

One of the reference technologies for the fuel cycle of fusion machines is Pd/Ag membranes. This technology is proposed to be implemented in tritium recovery systems because of their exclusive selectivity toward molecular hydrogen isotopes (Q = H, D, T). To perform scaling-up studies for the Tritium Extraction and Removal System of the European DEMOnstration fusion power reactor (DEMO) with a solid blanket, a one-dimensional simulation code was recently developed and successfully validated with experiments. This code relies on different operational (e.g., feed pressure and temperature), geometrical (e.g., permeator length), and membrane-intrinsic (e.g., Q2 permeability) parameters given as input. The main outcome is the Q2 permeation efficiency, defined as the Q2 permeate–to–feed flow ratio. Because of the low concentrations of Q2 expected at the He stream purging the solid blanket, the surface effects are expected to be important, decreasing the separation efficiency of the Pd/Ag permeators. In this paper the role of surface effects on the permeation efficiency is studied for a DEMO-relevant scenario (feeding mixture: HT/H2/He). Moreover, a sensitivity study is also given demonstrating the high impact of the permeation area, temperature, and feed pressure on the permeation efficiency of HT.