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The tokamak exhaust processing (TEP) system performs chemical separations on ITER fuel cycle process streams. TEP recovers hydrogen isotopes (Q2) from impurities such as argon, nitrogen, tr...

The ITER fuel cycle is composed of a tokamak and several systems that will support the preparation of fuel, the handling of exhaust gases, and the recycle of unused fuel back to the tokamak...

Zeolite membranes have been increasingly used for gas separation applications, where the most studied are the MFI and NaA types. Typically, NaA membranes are used for dehydration applications whereas MFI are used for the separation of large/small molecules (e.g., CO 2 /H 2 ). In the Tritium Laboratory Karlsruhe (TLK), both NaA and MFI-ZSM5 were selected for tritium processes where both molecular and oxidized tritiated species diluted in helium have to be processed in a continuous mode, as it is the case of the Tritium Extraction System of the Helium Cooled Pebble Bed breeding blanket. This paper presents and discusses the results obtained for the separation of a binary mixture of water vapour and helium, down to 0.2 vol%, where hydrogen replaces its radioactive counterpart. These experiments were performed in the temperature range between 25 and 100 °C. NaA demonstrates much better performances for feed concentrations ≥5 vol%, eventually reaching separation factors higher by a factor of 100 than that of MFI-ZSM5. However, the MFI-ZSM5 membrane must be selected instead when moisture levels are ≤1 vol%. In fact, at these conditions, the H 2 O/He separation factor for NaA barely exceeds 2, whereas it can be as high as 9.6 for MFI-ZSM5. Moreover, rather high permeances for H 2 O of around 1–2 μmol m −1 s −2 Pa −1 were measured for both membranes. In the case of helium, the permeances were highly dependent on the temperature for high water contents, reaching values as low as 1 × 10 −3 μmol m −1 s −2 Pa −1 for the NaA membrane.

Quantitative feasibility analysis of the tritium recovery efficiency from multiple columns of liquid PbLi droplets was conducted.Then a case study based on the HCLL specification was performed. Main concern was whether the experimentally obtained recovery efficiency from a column of droplets is applicable for the efficiency estimation from the multiple columns of droplets without any mutual degrading effects. To maintaining a safe side assumption, the tritium once released and reabsorbed on another droplet was considered to be not re-emitted while falling. By the analogy with the thermal radiation theory, the view factor which expresses the intersection ratio of radiation on another surface was applied for the estimation.The dependences on nozzle design parameters, such as nozzle pitch, number of nozzles, chamber wall clearance, and exhaust port design, were investigated. Case study results suggest that, by choosing well-suited parameters approximately 40% to 60% of the single column recovery effi...

Tritium behavior in a breeding blanket is a key design issue because of its impact on safety and fuel self-sufficiency best performance. Considering the difficulty in handling and its high cost, it is intended to prepare a simulation tool for tritium transport in the HCPB (Helium Cooled Pebble Bed) breeder blanket unit concept. The objective of this work is to evaluate tritium inventory inside several components of the breeder unit (pebble bed and purge gas) and its permeation into the coolant. Some simplified assumptions have been adopted and the results compared with others studies obtained by different simulation tools. Finally an example in which different experimental values of tritium residence time in ceramic breeder is presented with the purpose to observe the capability of the program to be integrated with experimental campaigns.

ZrCo is a well-known tritium storage material and has been studied intensively in the literature. The most interesting properties with regards to the thermodynamics of the ZrCo-H system are the very low H2 partial pressure in equilibrium with ZrCoH3 at room temperature and the ease to reach sufficiently high temperature to completely release the stored H2. These properties motivate also to use ZrCo not as a simple storage, but rather as a concentrator of hydrogen isotopologues from inert gases like He. With such function, ZrCo getter beds are the reference solution adopted in the conceptual design of the tritium extraction system of the European Test Blanket Modules (TBM) to replace the cryogenic molecular sieve bed previously proposed. An experimental campaign was carried out on ZrCo in order to consolidate this choice. The results confirmed that ZrCo performs well as getter material but only substantially below the maximum loading capacity. They revealed that the dynamic thermo-mechanical respon...

Experiments are still necessary to consolidate the processes retained for the Tritium Extraction Systems of the European ITER Test Blanket Modules (TBM). A PERMCAT reactor combines a catalyst promoting isotope exchange reactions and a Pd/Ag membrane allowing tritium recovery from complex gaseous mixtures containing tritium in different chemical forms. Originally developed for the Tokamak Exhaust Processing, the PERMCAT process is also candidate to detritiate the water arising from an adsorption column installed in the TBM ancillary systems. We discuss the results of an extensive experimental campaign using a PERMCAT reactor to process Q2O containing impurity gas mixtures at high flow rates. Two different experimental configurations were studied, namely PERMCAT stand-alone, and PERMCAT in combination with a zeolite molecular sieve bed (MSB, previously loaded with Q2O) under regeneration. On the one hand, many expected behaviors were observed, such as the key influence of the swamping ratio (ratio between the impurity and purge side flow rates) on the decontamination factors. On the other hand, some rather unexpected effects tend to show that the limiting phenomena of such specific operation of PERMCAT reactors (at high flow rates, thus short residence time) lies on the kinetics of the isotope exchange reactions.

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.

Zeolite membranes were identified to be used in the future fusion reactors to separate traces of tritiated species from large helium flows. Due to the lack of data in the literature, separation experiments with helium, hydrogen and water vapor have been performed in the Tritium Laboratory Karlsruhe (Germany). The obtained results shall be used to predict performances at reactor scale. Therefore, a simulation tool of a permeator has been developed to perform calculations ab-initio of its performance at several experimental conditions. In the present work, numerical results for the binary mixture hydrogen/helium at different concentrations are compared with the experimental results obtained with a zeolite MFI-ZSM5 tubular membrane, where separation factors below 2 were measured for different cuts. Very good agreement between experiments and simulations was observed. This agreement gives confidence to integrate this numerical tool into a membrane cascade simulator, in development in the TLK, toward the modeling of the zeolite membranes’ performances at reactor scale.

Groundwater contaminated with geogenic arsenic (As) is frequently used as drinking water in Burkina Faso, despite adverse health effects. This study focused on testing low-cost filter systems based on zero-valent iron (ZVI), which have not yet been explored in West Africa for As removal. The active ZVI bed was constructed using small-sized iron nails, embedded between sand layers. Household filters were tested for nine months in a remote village relying on tube well water with As concentrations of 400–1350 μg/L. Daily filtered volumes were 40–60 L, with flow rates of ~10 L/h. In parallel, downscaled laboratory filter columns were run to find the best set-up for optimal As removal, with special attention given to the influence of input pH, flow rate and water/nail contact time. Arsenic removal efficiencies in the field were 60–80% in the first six months of operation. The laboratory experiments revealed that trapped air in the nail layer greatly lowered As removal due to preferential flow and decreased water/nail contact time. Measures taken to avoid trapped air led to a partial improvement in the field filters, but effluent As remained >50 μg/L. Similar structural modifications were however very successful in the laboratory columns, where As removal efficiencies were consistently >95% and effluent concentrations frequently 1000 μg/L. A constantly saturated nail bed and careful flow control is necessary for optimal As removal. Slow flow and longer pauses between filtrations are important for sufficient contact times and for transformation of brown amorphous Fe-hydroxides to dense magnetite with incorporated As(V). This preliminary study has shown that nail-based filters have the potential to achieve As removal >90% in a field context if conditions (filter bed saturation, flow rate, pauses between filtrations) are well controlled., Science of The Total Environment, 737, ISSN:0048-9697, ISSN:1879-1026

This paper describes the main acceptance criteria and required acceptance tests for the components of the six Test Blanket Systems to be installed and operated in ITER. It summarizes the guide-line toward the establishment of detailed test plans for the TBS, starting from the end-product at the ITER Members factories, and to generally define the type of tests that have to be performed on the ITER site after shipment, and/or prior to the systems final commissioning phase.

Highly tritiated water (HTW) is produced in fusion machines and one of the promising technologies to process it is isotopic exchange. 3 kinds of Pt-catalyzed zeolite (13X-APG, CBV-100-CY and HiSiv-1000) were tested as candidates for isotopic exchange of highly tritiated water (HTW), and CBV-100-CY (Na-Y type with a SiO{sub 2}/Al{sub 2}O{sub 3} ratio of ∼ 5.0) shows the best performance. Small-scale tritium testing indicates that this method is efficient for reaching an exchange factor (EF) of 100. Full-scale non-tritium testing implies that an EF of 300 can be achieved in 24 hours of operation if a temperature gradient is applied along the column. For the isotopic exchange, deuterium recycled from the Isotope Separation System (deuterium with 1% T and/or 200 ppm T) should be employed, and the tritiated water regenerated from the Pt-catalyzed zeolite bed after isotopic exchange should be transferred to Water Detritiation System (WDS) for further processing.

The PETRA facility is the first installation in which experiments with tritium were carried out at the Tritium Laboratory Karlsruhe. After completion of two main experimental programs, the decommis...

The CAPER facility at TLK originally devoted to R&D on tokamak exhaust processing has been significantly upgraded over the last years. Beside new R&D on highly tritiated water, CAPER is presently l...

The CAPER facility of the Tritium Laboratory Karlsruhe has demonstrated the technology for the tokamak exhaust processing. CAPER has been significantly upgraded to pursue research/development programs towards highly tritiated water (HTW) handling and processing. The preliminary tests using a metal oxide reactor producing HTW afterward de-tritiated with PERMCAT were successful. In a later stage, a micro-channel catalytic reactor was installed in view of long term research program on HTW. The integration of this new system in CAPER was carried out along with a careful safety analysis due to high risk associated with such experiments. First experiments using the μ-CCR were performed trouble free, and HTW up to 360 kCi/kg was produced at a rate of 0.5 g/h. Such HTW was collected into a platinum zeolite bed (2 g of HTW for 20 g of Pt-zeolite), and in-situ detritiation was performed via isotopic exchange with deuterium. These first experimental results with tritium confirmed the potential for the capture and exchange method to be used for HTW in ITER. (authors)

Membrane separation by zeolite membranes has been proposed as a pre-concentration stage for the tritium extraction from the purge helium of the breeding blanket combined with a final recovery by th...

Tritium Extraction Systems (TES) in future nuclear fusion power plants will recover tritium (radioactive hydrogen isotope) from the outlet stream of the breeder blanket, where tritium will be produced and purged out by a high flow rate of helium. This work presents an experimental assessment of the performance of a MFI zeolite hollow fibres membrane for the tritium removal and pre-concentration in the first stage of TES. A set-up was built for the permeation experiments of hydrogen and helium single gases and the separation experiments of binary mixtures of those gases. Numerous assays were performed for binary mixtures of H2 and He under various operating conditions (feed and permeate pressures, temperature, and feed composition) to assess the membrane separation performance. Optimal separations were achieved at the lowest temperatures (around room temperature) and the lowest hydrogen molar fraction (0.22). H2 and He permeances ranged 0.70–2.3 and 0.43–0.74 μmol m−2 Pa−1 s−1, respectively, corresponding to membrane selectivities of H2 over He between 1.2 and 4.0. By regulating the feed pressure, hydrogen recovery factors around 50% were attained without compromising the maximum enrichment factor of 1.5. These preliminary data enable the design of a multi-stage separation process complying with the strict requirements for the tritium recovery and enrichment in a nuclear fusion power plant.

While the tritium technology for the inner DT fuel cycle of fusion reactors shall be demonstrated in ITER, the tritium management in the breeder blanket remains very challenging. Most of the process options rely on ad(b)sorption/desorption cycles, using dedicated packed beds to handle separately the molecular and oxide forms of tritium. This approach seems satisfactory for ITER, but seems difficult to scale up to DEMO. The alternative use of a catalytic membrane reactor in combination with inorganic membranes would simplify and improve the overall tritium management. Zeolite membranes should enable in a single step the pre-concentration of all tritiated species. This tritium enriched stream could be afterwards processed using PERMCAT (catalytic Pd-based membrane reactor) to finally recover the tritium in its pure molecular form. This paper discusses at the conceptual level such approach. The latest experimental results on zeolite membrane and multi-tube PERMCAT reactor are presented. Next R&D activities for technical scale demonstrations and refined simulation tools are proposed to finally estimate the sizes of the components to be operated in ITER and DEMO.

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. © 2013 Elsevier B.V. All rights reserved.

International audience; 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 integrated design and a RandD program 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.

In future DT fusion machines, several events will generate highly tritiated water (HTW). Among potential techniques for HTW processing, isotopic swamping in a catalytic membrane reactor (PERMCAT) appears promising. The experimental demonstration of PERMCAT for HTW processing has started in the CAPER facility at the Tritium Laboratory of Karlsruhe (TLK). Without any HTW source, such water has to be produced on purpose. Catalytic HT oxidation would ensure clean operation but could be critical for operation due to possible occurrence of explosive mixture. A tritium compatible micro-channel catalytic reactor (μCCR) has been designed and manufactured to produce up to 10 mL min−1 of HTW with very high specific tritium activity (stoichiometric DTO: 5.2 × 1016 Bq kg−1). Prior to its integration in CAPER for tritium operation, this reactor has been commissioned at different feed flow rates, gas composition (air or Helium), and temperature. The results demonstrate the good performances of the μCCR in producing water. The combination of the μCCR with the O2 sensor represents a reliable system able to produce HTW in a safe way and without radioactive waste. Accordingly, the CAPER facility can be upgrade in order to continue the R&D activity on HTW processing with PERMCAT.

Tritium handling facilities use molecular sieve beds (MSB) to collect and recover tritiated water. After reaching the capacity limit of the MSB, the water is desorbed and decontaminated in a water detritiation system (WDS). In the case of highly tritiated water (HTW) absorbed into a MSB, an inherent safe option for processing is necessary due to the HTW specific properties. Ideally, HTW should be processed immediately in a continuous mode. With this in consideration, the water desorption process from a zeolite bed was developed and optimized in a dedicated non active facility. The results of this experiments were applied into the regeneration of a MSB previously loaded with HTW containing an activity of 1.9 × 1014 Bq kg−1. The water was desorbed, by step increasing the temperature bed and fed by helium carrier gas into the PERMCAT for detritiation and tritium recovery. The processed water was collected in a dry MSB downstream of the PERMCAT. These initial studies successfully demonstrate the viability of the process. The obtained results of the preliminary study and the subsequent tests with tritium, will provide useful information for the design of tritium processes relying on MSB, such as the water processing foreseen for the test blanket modules in ITER.

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 80 bar 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.

Safe, reliable, and efficient tritium management in the breeder blanket will have to face unprecedented technological challenges. Beside the efficiency for tritium recovery from the breeder blanket (Tritium Extraction (TES) and Coolant Purification Systems (CPS)), the accuracy for tritium tracking between the inner and the outer fuel cycle must also be demonstrated. This paper focuses on the recent R&D carried out at the Tritium Laboratory Karlsruhe to tackle these issues. For ITER, the recently consolidated TES and CPS designs comprise adsorption columns and getter beds operated in semi-continuous mode. Different approaches for the tritium accountancy stage (TAS) have been evaluated. Balancing static (batch-wise gas collection at the TBM outlets and the tritium plant) or dynamic (in/on-line) approaches with respect to the expected analytical performances and integration issues, the first conceptual design of the TAS for EU TBMs is presented. For DEMO, the overall strategy for tritium recovery and tracking has been revisited. The necessity for on-line real-time tritium accountancy and improved process efficiency suggest the use of continuous processes such as permeator and catalytic membrane reactor. The main benefits combining the PERMCAT process with advanced membranes is discussed with respect to process improvements and facilitated accountancy using spectroscopic methods.

In view of future fusion rectors fueled by deuterium–tritium mixtures, highly tritiated water (HTW) of up to 5.2·1016 Bq kg−1 will be produced, during routine operation and scenarios as an accidental release of tritium into a glove box. Also in the solid breeder blanket concept, a non-negligible fraction of the tritium produced will be in the tritiated water fraction. To decontaminate HTW the PERMCAT using isotope swamping in a Pd/Ag membrane reactor has been identified as a robust and reliable solution. In order to investigate the decontamination of HTW at flow rates relevant for future fusion power plants, a technical scale, fully tritium compatible PERMCAT consisting of a bundle of finger-type membranes inserted in a single catalyst bed was developed. Nevertheless, it represents only one part of a PERMCAT cascade necessary to achieve the required performance to process HTW on technical scale. By improving the existing PERMCAT geometry using experimental data obtained from isotopic exchange between D2O and H2, the performance of the existing PERMCAT reactor was optimised. Based on the optimised geometry a new fully tritium compatible technical scale PERMCAT cascade comprising of two PERMCAT reactors in series was designed, manufactured and commissioned as presented in this paper.

The Coolant Purification System (CPS) together with the Tritium Extraction System (TES) and Helium Cooling System (HCS) are the principal auxiliary circuits of HCLL and HCPB TBMs to be tested in IT...

The CAPER facility operated at the Tritium Laboratory Karlsruhe for the demonstration of the tokamak exhaust processing system comprises a PERMCAT reactor as final clean-up stage. CAPER has been up...

The plasma exhaust gas from a fusion machine needs to be processed with very high tritium removal efficiency. Such a high decontamination factor is achieved by multistage processes. The last stage of the three-step CAPER process developed at the Tritium Laboratory Karlsruhe (TLK) is based on a so-called permeator catalyst (PERMCAT) reactor. The PERMCAT principle is based on isotopic swamping in a counter current mode. To support experiments, dynamic process simulation is indispensable for process design, analysis, and operation. A first numerical calculation code has been developed for the process simulation of PERMCAT. The numerical model developed was based on counter current one-dimensional plug flows partitioned with a membrane. In this study, the axial dispersion was newly taken into consideration to the numerical simulation code and its effect was examined. Improvement of the numerical simulation code considering a 2-dimensional model was developed and test calculations were carried out.

Tritium used as fuel in future fusion machines will be produced within the breeder blanket. The tritium extraction system recovers the tritium to be routed into the inner-fuel cycle of the machine. Accurate and precise tritium accountancy between both systems is mandatory to ensure a reliable operation. Handling in the blanket huge helium flow rates containing tritium as traces in molecular and oxide forms is challenging both for the process and the accountancy. Alternative tritium processes based on combinations of membranes and membrane reactors are proposed to facilitate the tritium management. The PERMCAT process is based on counter-current isotope swamping in a palladium membrane reactor. It allows recovering tritium efficiently from any chemical species. It produces a pure hydrogen stream enriched in tritium of advantage for integration upstream of the accountancy stage. A pre-separation and pre-concentration stage using new zeolite membranes has been studied to optimize the whole process. Such a combination could improve the tritium processes and facilitate accountancy in DEMO.

At the Tritium Laboratory Karlsruhe stainless steel cross-piece ionization chambers have been used to measure the activity concentration of tritiated gases in experiments and processes for more than 10 years. New chambers with an optimized design in terms of the effective chamber volume were produced. Furthermore, they were gold and copper plated to determine the influence of the coatings on the signal and on the memory effect. A new chamber of the old design was built for direct comparison of the signals. The chambers were characterized with different helium-tritium mixtures in 8 runs and the ionisation current as a function of the static gas pressure was measured. When comparing the three new chambers, the gold chamber always showed the highest current, followed by the copper chamber. After exposing the chambers to ~13,100 TBq/m3, the memory effect was investigated by using a similar gas mixture of the earlier runs with ~1,500 TBq/m3. The gold chamber showed the highest memory effect, the copper...

In the frame of the design and development of the Water Detritiation System (WDS) for ITER, the TRENTA facility was installed at the Tritium Laboratory Karlsruhe (TLK). This facility allows processing up to 1.7 kg h−1 of tritiated water, it is however restricted to a maximum activity concentration of 3.7 × 1011 Bq kg−1. Because highly tritiated water (HTW) still is an issue for fusion technology, it is necessary to develop an additional process to decontaminate the highly tritiated part of the water waste stream. In order to provide a viable solution to process tritiated water in excess of 1013 Bq kg−1 an inactive experimental study was carried out to substantiate the design of an experimental facility based on a PERMCAT type catalytic palladium membrane reactor. A feasibility study was made to provide an outline design for an experimental plant to be used for parametric investigations with tritiated water of up to 5.2 × 1016 Bq kg−1. In parallel to experiments with a large technical PERMCAT using D2O and protium, a design study was made to define process steps and interfaces relevant for a technical scale experiment using a PERMCAT to decontaminate HTW at TLK.

The PERMCAT membrane reactor is a coaxial combination of a Pd/Ag permeator membrane and a catalyst bed. This device has been proposed for processing fusion reactor plasma exhaust gas. A stream containing tritium (up to 1% of tritium in different chemical forms such as water, methane or molecular hydrogen) is decontaminated in the PERMCAT by counter-current isotopic swamping with protium. Different mechanical designs of the membrane reactor have been proposed to improve robustness and lifetime. The ENEA membrane reactor uses a permeator tube with a length of about 500 mm produced via cold-rolling and diffusion welding of Pd/Ag thin foils: two stainless steel pre-tensioned bellows have been applied to the Pd/Ag tube in order to avoid any significant compressive and bending stresses due to the permeator tube elongation consequent to the hydrogen uptake. An experimental test campaign has been performed using this reactor in order to assess the influence of different operating parameters and to evaluate the overall performance (decontamination factor). Tests have been carried out on two reactor prototypes: a defect-free membrane with complete (infinite) hydrogen selectivity and not perm-selective membrane. In this last case, the study has been aimed at verifying the behaviour of the PERMCAT devices under non-normal (accidental) conditions in the view of providing information for future safety analysis. The paper will present the specific mechanical design and the experimental results of tests based on isotopic exchange between H2O and D2.

The PERMCAT process developed for the final clean-up stage of the Tokamak Exhaust Processing systems of the ITER tritium plant combines a catalytic reactor and a Pd/Ag permeator in a single component. A first generation technical PERMCAT has been successfully operated as part of the CAPER experiment at the Tritium Laboratory Karlsruhe for several years. Various alternative PERMCAT mechanical designs were proposed and studied on small-scale prototypes. An upgraded technical PERMCAT reactor was designed, manufactured and commissioned with deuterium. A parallel arrangement of finger-type membranes inserted in a single catalyst bed design was chosen to simplify the geometry and the manufacturing while improving the robustness of the reactor. The component has been designed and manufactured to be fully tritium compatible and also fully compatible with both process and electrical connections of the previous PERMCAT to be replaced. The new PERMCAT mechanical design is more compact and easy to manufacture. This PERMCAT reactor was submitted to functional tests and experiments based on isotopic exchanges between H2O and D2 to measure the processing performances. The first experimental results show decontamination factors versus flow rates better than all previously measured.

European test blanket modules (TBM) for both European blanket concepts, namely the Helium Cooled Lithium Lead (HCLL) and the Helium Cooled Pebble Bed (HCPB), and their auxiliary systems have been designed in the last years. This paper presents the experimental facilities available in Europe for the qualification tests of European TBMs. In particular, devices designed to analyze specific technological issues, for example corrosion under strong magnetic fields or tritium management and extraction, as well as large scale facility to perform integral tests are described. A preliminary overview on the qualification strategy is presented and discussed.

The HCPB (Helium Cooled Pebble Bed) and HCLL (Helium Cooled Lithium Lead) Test Blanket Modules (TBMs), developed in EU to be tested in ITER, adopt helium at 80 bar as primary coolant. This paper contains a conceptual design of the TBMs Coolant Purification System (CPS) based on the need to remove permeated tritium and gas impurities. The following steps have been considered: identification of CPS design requirements; review of the purification systems developed for Helium Cooled Fission Reactors and proposed for DEMO Fusion Reactor; selection of the most promising technologies for CPS; indications on instrumentation and procedures for tritium balance. The proposed solution is a three-stage process constituted by an oxidiser to convert Q 2 and CO to Q 2 O and CO 2 , an adsorption step, performed on molecular sieve at room temperature to remove Q 2 O and CO 2 , and a final step performed on a heated getter to remove residual impurities.

After 8 years of operation at the CAPER facility at the Tritium Laboratory Karlsruhe, a permeator used to separate hydrogen species from processed gases ceased operation due to multiple heater failures. This was subjected to post service examination to find the cause of the failures. This paper describe the methods used to locate the failures in the heaters and the likely cause. It was also necessary to determine the tritium inventory embedded in the structure for safe disposal. Destructive examination, adapted from a full combustion technique, was used on sections of the permeator. A fine black powder deposit, presumed to be mostly carbon, coated the surfaces of the inlet section of the feed side. This powder contained nearly half of the tritium within the permeator. The likely source of the powder and the consequences for the operation and eventual decommissioning of the ITER Tritium Plant are discussed. A failed turbomolecular pump from CAPER was also examined. There was evidence of wear on the emergency support bearing, but more importantly, when the pump internals were exposed to the glove box atmosphere (dry air) large quantities of tritium were rapidly released, this despite the isotopic swamping before removal from the CAPER glovemore » box. Significant uptake of tritium in electrical insulation was also found. (authors)« less

The PERMCAT process, chosen for the final clean-up stage of the Tritium Exhaust Processing system in ITER, directly combines a Pd/Ag membrane and a catalyst bed for the detritiation of gaseous mixtures containing molecular and chemically bound tritium. Upgraded PERMCAT mechanical designs have been proposed to both increase the robustness and simplify the design of the reactor. One uses a special corrugated Pd/Ag membrane able to withstand change in length of the membrane during both normal operation and in the case of off-normal events. Based on this design, an upgraded PERMCAT reactor has been produced at FZK and successfully tested at TLK with ITER relevant tritiated gaseous mixtures using the CAPER facility. (authors)

The Tritium Laboratory Karlsruhe (TLK) has been designed to handle relevant amounts of tritium for the development of tritium technology for fusion reactors. This paper describes the tritium technology development and experience gained during the upgrade of facilities, interventions, replacement of failed components and operation of the TLK since its commissioning with tritium in 1994. (authors)

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.

An inverse isotope effect is observed for both the solubility and the diffusivity of hydrogen in palladium, i.e., the lightest isotope has the highest solubility and lowest diffusivity. For permeation of hydrogen isotopes through palladium alloy membranes when diffusion is the rate determining step tritium is the least permeating isotope. Since the absolute values of hydrogen, deuterium and tritium permeabilities in palladium differ considerably, the effect could be employed as a basis for technical isotope separation. Using an experimentally verified mathematical model, a parametric computational study was performed to show the net isotope effects in permeate and bleed flows when feeding a technical permeator with various hydrogen isotope mixtures under different feed and permeate pressures. The technical feasibility of hydrogen isotope permeation as a method for separation is discussed with regard to the necessary process control for either a single permeator or a cascade.

Tritium processing technologies of the two European Test Blanket Systems (TBS), HCLL (Helium Cooled Lithium Lead) and HCPB (Helium Cooled Pebble Bed), play an essential role in meeting the main objectives of the TBS experimental campaign in ITER. The compliancy with the ITER interface requirements, in terms of space availability, service fluids, limits on tritium release, constraints on maintenance, is driving the design of the TBS tritium processing systems. Other requirements come from the characteristics of the relevant TBM and the scientific programme that has to be developed and implemented. This paper identifies the main requirements for the design of the TBS tritium systems and equipment and, at the same time, provides an updated overview on the current design status, mainly focusing onto the tritium extractor from Pb-16Li and TBS tritium accountancy. Considerations are also given on the possible extrapolation to DEMO breeding blanket. © 2015, American Nuclear Society. All rights reserved.

Zeolites are known as tritium compatible inorganic materials widely used in packed beds as driers in detritiation systems and are also suggested for tritium removal from helium at cryogenic temperature. The Tritium Laboratory Karlsruhe (TLK) proposed a new fully continuous approach for tritium extraction from the solid breeding blanket of fusion machines that improves the overall tritium management and minimizes both the tritium inventory and processing time. It is based on membrane permeation as a pre-concentration stage upstream of a final tritium recovery stage using a catalytic Pd-based membrane reactor. Zeolite membranes were identified as the most promising candidates for the pre-concentration stage. In the present work the tubular zeolite MFI membrane provided by the Institute for Ceramic Technologies and Systems (IKTS, Hermsdorf, Germany) is studied to consolidate the proposed approach. The permeation measurements for single gases hydrogen (replacing radioactive tritium) and helium, for binary mixtures H2/He and H2O/He at different concentrations and temperatures are presented. The tested membrane demonstrates a high performance, almost independent from the inlet composition in the case of a gaseous mixture, while the transport in the presence of water vapour is strongly related to the temperature of the mixture and component concentrations.

A traditional helium leak detector mass spectrometer is applied in analytical chemistry. We report its straightforward use for the precise quantification of 3He and 4He concentrations at a sub-ppb level. Repetitive external calibrations with certified gases demonstrate long-term stability, reproducibility of 3% at the ppb range, and linearity versus helium concentration over 5 orders of magnitude. The mass spectrometric determination of 3He takes into account a H3+ contribution and has a 3He detection limit of 2 ppb. Since 4He induces an attenuation effect of the mass 3 signal, this apparatus is only suitable for the measurement of 3He/4He ratio greater than 10 Ra (atmospheric ratio Ra = 1.38 x 10(-6)). This mass spectrometric technique has found novel application in the determination of tritium concentrations with an original 3He ingrowth method. Here, samples are first purified cryogenically onto activated charcoal to eliminate tritium interfering during the 3He mass spectrometric determination step. Helium is also preconcentrated, and 3He is routinely determined around 0.05 ppb with a 20% relative uncertainty. This 3He measurement technique has been successfully applied in the field of tritiated waste drums as a nondestructive method. The total amount of tritium present in these 200-L drums can be determined as low as approximately 1 GBq.

This paper presents a method to quantify the tritium activity in a drum by measurement of its helium-3 leak. A model of helium exchanges with the atmosphere has been developed. It takes into account the diffusion phenomena and the influence of atmospheric pressure changes. The validation has been achieved with a pilot and the comparison between theoretical and experimental data has highlighted a very good agreement. Drum's helium-3 leak equilibrates after six months and then equals the helium-3 production in the drum and so to the total tritium activity. The measurement technique is also described. It's based on quantitative helium trace level determinations with an adapted leak detector. After a drum's confinement period of 5 hours and a cryogenic treatment of gaseous samples, the method allows to detect a 5 GBq activity of tritium in a drum.

The availability of calibrated gas mixtures in a tritium plant is an important precondition for correct analysis of unknown gas samples because a quantitative analysis with a gas chromatograph (GC) requires frequent calibration checks. To calibrate a GC for tritium, certified gas mixtures with different tritium concentrations are necessary. These mixtures can be produced directly in a laboratory. In this case an independent method is needed to determine the tritium concentration in the mixture with high accuracy. Calorimetry is the method of choice because its accuracy is better than 1% and it is non-destructive method. Using the closed tritium loop of the Tritium Laboratory Karlsruhe (TLK), different mixtures containing tritium in the range 1–100% have been produced and measured with one of the in-house calorimeters. After that the GC of the CAPER facility at TLK has been calibrated several times with an uncertainty of a few percents.

The next generation of fusion power plants (DEMO) should rely on a breeding blanket (BB) able to efficiently convert the neutrons kinetic energy into heat, 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 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 the $K-{S}$ values, this paper provides a parametric analysis for the most influencing parameters related to tritium migration in helium-cooled and water-cooled lead-lithium DEMO BBs. This analysis has been performed using the computational code FUS-TPC for several operative scenarios and considering the different KS values. This paper 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. © 2014 IEEE.

The design of ITER tritium processing systems should benefit from experimental data and process validation on experimental facilities that are ITER relevant. TriPla-CA, which is a consortium of four EU laboratories aiming to support tritium relevant activities for ITER design is carrying out R&D and design for ITER Tritium Plant. Several rigs and experimental facilities have been enhanced and developed in order to explore a wide range of envisaged scenarios of some ITER Tritium Plant systems such as water detritiation system (WDS), isotope separation system and highly tritiated water processing. Beside the experimental activities, the enhancement of the relevant software for simulation and design of various tritium processing systems is ongoing. The main achievements concerning the R&D, the mechanical design including the seismic calculations of some ITER WDS components and the main expertize and the hardware available inside the consortium are presented. © 2014 Elsevier B.V.

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. © 2014 Elsevier B.V.

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). © 1973-2012 IEEE.

Tritium permeation through Breeding Blanket 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 in order to predict the released amount of tritium during the plant operation. Unfortunately, tritium assessments are often affected by several uncertainties implying very important modelling and parametric issues. In this study the main permeation issues are identified and possible solutions are discussed to face the modelling issues and the parametric uncertainties affecting the T migration assessments for the two DEMO helium-cooled breeding blankets, i.e.: 1) Helium-Cooled Pebble Beds (HCPB) and 2) Helium-Cooled Lithium-Lead (HCLL). For these two helium-cooled blanket concepts various tritium migration analyses will be carried out by means of the computational tool FUS-TPC in order 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). © 2013 IEEE.