European Test Blanket Modules Research Articles

Tritium transport modelling in Europe: Current status, multi-physics integration and proposal of material property data

The capability to predict tritium retention and permeation by accurate and reliable tritium transport models for the breeding blanket design development is one of the most relevant objectives of the European Test Blanket Module (TBM) Programme. This also represents a key aspect for the development and licensing of next-future fusion systems and power plants, including the European DEMO. Tritium analyses can be performed using either system-level models to calculate global blanket and the related ancillary system performances or detailed 3D models for spatially limited areas and critical regions: the choice between these two modelling approaches depend on the objective of each analyse. When looking at the accuracy of tritium transport simulations, two main aspects have to be considered: the first one is the completeness of the description of the physical phenomena participating to the tritium transport dynamics; the second one is the reliability of the tritium transport material property values used in the simulation activities.In this context, this paper has the objective to carry out a critical analysis on the status of the integration of the most relevant multi-physics elements, providing recommendations on the approach to follow, with a particular focus on the system-level codes. Moreover, a set of data and correlations on tritium transport materials properties is proposed as a first step towards a material property database to be shared within the breeding blanket community.

An integral methodological framework for the thermo-mechanical analysis and structural integrity assessment of the European TBM sets

Two European Test Blanket Modules (TBM) concepts are being developed by F4E, the Water-Cooled Lithium Lead (WCLL) and the Helium-Cooled Ceramic Pebbles concept (HCCP, in collaboration with ITER KOREA). For both concepts, the ESTEYCO Mechanics team is responsible for designing and analyzing the TBM-box, which serves as the plasma-facing module for tritium breeding, and the TBM Shield in the WCLL concept. The design-by-analysis activities for these TBM-Sets are particularly challenging due to several factors. These include diverse and complex loads such as thermal, electromagnetic, pressure, and seismic loads, intricate failure modes to be assessed, the use of a new structural material (EUROFER 97) with limited available information, and the need to work within tight compliance margins in a restricted design space. Over the years, ESTEYCO has developed advanced methodologies to virtually test various design solutions systematically, coherently, and efficiently, aligning with the applicable nuclear design code (RCC-MRx). This paper provides an overview of key developments in simulating the TBM-Set under a comprehensive set of loads and subsequently assessing its structural integrity in an exceptionally detailed and comprehensive manner. The authors have created novel automated tools that encompass the entire process. These tools address tasks like distributing nuclear thermal loads, simulating cooling system behavior under operational conditions, characterizing and distributing electromagnetic loads, implementing code assessment rules for numerous evaluation points, and developing robust visualization tools for result analysis and validation.

An update on the latest and most relevant improvements achieved in the European helium-cooled ceramic breeder TBM design

The European TBM (Test Blanket Module) program under development will test two alternative concepts for tritium generation and extraction in ITER: a helium cooled / ceramic pebble breeder concept and a water cooled / lithium lead breeder concept.The former implements Advanced Ceramic Breeder pebbles that act as tritium breeding material, and Beryllium pebbles as neutron multiplier. A Reduced Activation Ferritic Martensitic steel, the EUROFER-97 (X10CrWVTa9-1), is used as structural material and pressurized Helium technology is implemented for efficient heat extraction (300-500°C, 8 MPa).The design of the TBM box involves challenges that must be tackled with a holistic approach to simultaneously consider a variety of aspects. Some of these challenges include: the thermal performance and the structural integrity of the TBM under ITER normal operations and accidental conditions, the implementation of pressurized equipment regulation(s), the convergence with manufacturing activities being developed in parallel and the high congestion in the reduced TBM volume –especially when considering instrumentation integration– without affecting too much the tritium breeding capability. It is also key to develop advanced methodologies based on design by analysis to virtually test alternative design solutions in an agile manner. These should be linked to the applicable nuclear design code (RCC-MRx) to consolidate design choices. In this way, a traceable evolution of the TBM design process can be built where alternatives are tested, compared and either discarded or selected in a consistent and comprehensive framework.This paper presents an overview of the very significant progress achieved in the design of the helium cooled concept TBM over the last two years, a period that corresponds to the first steps of the preliminary design phase towards the Preliminary Design Review (PDR) gate. This includes a significant change in the design of the back manifold area and the connection with the rest of the TBM-Set that solves long-standing historical problems inherited from pre-conceptual design stage. An update on the latest and most relevant methodological developments achieved is also provided in this paper, since these have been key to progress with the design evolutions in a sound and traceable manner. To end, the paper also presents some examples of the instrumentation definition and its complex integration within and outside the TBM box.

Practical considerations in developing nuclear detectors for tokamak harsh environments

Fusion tokamaks (FT) and hybrid fusion-fission reactors (HFFR) present harsh working conditions characterized by intense neutron and gamma fluxes (>1012 cm−2 s−1), high working temperatures (up to 600 °C) and corrosive environment. The breeding blanket region (BB) of these plants are resulting very hostile to the detectors used to monitor/measure fundamental nuclear parameters such as neutron/gamma fluxes and energy spectra, and tritium production. Presently no detectors are ready for being hosted in the harsh environment of the BB and R&D activity is needed to develop and test the candidate detectors. Some important lessons can be learned from past activities carried out in the EU and devoted to studying and realizing nuclear detector prototypes for the European Test Blanket Modules (TBM) of ITER. Amongst the other, these studies pointed out the need for intense neutron fields and calibration facilities closely reproducing the expected working environments to be used for reliable testing and calibration of the prototypes.Accurate simulation by Monte Carlo technique of the proposed detectors allows to mimic and foresee the response and performances of the detectors pointing out several fundamental and critical aspects on the physical response of the detector so helping in understanding the detectors response. This can help in selecting the best performing detector. The selection is based upon a multi-step procedure.The lesson learned for ITER-TBM can be helpful to study and develop nuclear detectors to be used in HFFR reactors and in next fusion machines like DEMO this because, despite the difference, the ITER-TBMs and the BB of fusion devices and HFFR reactors experience a number of similarities in terms of radiation level, temperature and nuclear quantities to be measured.In this paper, after discussing the requirements to be fulfilled by the nuclear detectors that must operate in the harsh environments we will discuss an example of detector development by considering the case of a self-power neutron detector (SPND) with chromium emitter studied and developed for ITER-TBM. The detailed Monte Carlo analysis is also reported and the many issues not yet solved are highlighted and the possible follow up to HFFR instrumentation discussed.

Thermo-hydraulic analysis of PbLi ancillary system of WCLL TBM undergoing in-box LOCA

The European liquid Test Blanket Module (TBM) concept named Water Cooled Lithium Lead makes use of the liquid metal alloy PbLi as tritium breeder/neutron multiplier and water as coolant. A dedicated PbLi ancillary system feeds the detritiated liquid metal alloy to the TBM and returns the PbLi enriched with tritium. The object of this study is developing a numerical model able to simulate the thermo-hydraulic behavior of the PbLi ancillary system under both normal operation and accidental scenario, with the aim to investigate the propagation of pressure waves generated by an in-box LOCA. The accident consists of a tube rupture of the coolant circuit inside the TBM, leading to the interaction between PbLi and water. As a consequence, a pressure increase occurs inside the breeder zone. The accident might affect the ancillary systems due to the propagation of the pressure waves generated inside the TBM. The PbLi loop is modelled by means of the thermo-hydraulics system code RELAP5/Mod3.3 and an instantaneous pressure peak is applied inside the control volume simulating the TBM soon after reaching the steady state conditions inside the whole loop. The propagation of the pressure wave inside the circuit and the hydraulic behavior of the individual components is studied. The present paper represents a first step towards the investigation of the PbLi system and its safety features with the aim to support the R&D activities carried out in the framework of the ITER project.

Fabrication technologies implemented for the European test blanket modules: Status for HCPB and WCLL

Abstract Within the framework of the European fusion strategy, two tritium Breeder Blankets concepts are developed to be tested in ITER as Test Blanket Modules (TBM): the Water-Cooled Lithium-Lead (WCLL) which uses the liquid Pb-16Li as a breeder, neutron multiplier and tritium carrier, and the Helium-Cooled Pebble-Bed (HCPB) with lithiated ceramic and beryllium pebbles as breeder and neutron multiplier, respectively, and helium purge gas as tritium carrier. Both concepts consider as structural material reduced activation ferritic martensitic steel, the EUROFER97 (X10CrWVTa9-1). Pressurized water (15.5 MPa, 295−328 °C for WCLL) and pressurized helium (8 MPa, 300−500 °C for HCPB) are foreseen for heat removal. The TBM is constituted of a BOX (side caps, first wall), stiffened by stiffening plates (SP) and closed on its back, in the manifold area, with back plates with passing-through elements, like stiffening/tie rods, SP nozzles and inlet/outlet water/He pipes. Internally, breeder units, delimited by SP, are cooled-down by cooling plates for the HCPB and double-wall tubes for the WCLL. This paper describes manufacturing technologies for the TBM and deals with the definition of the preliminary Welding Procedure Specifications (pWPS) and feasibility mock-ups for the manifold area and TBM BOX assembly. The transfer and validation of the pWPS developed for the TBM, from P91 to EUROFER97 steel, is demonstrated. Consolidation of the pWPS for the TBM BOX assembly made with HIP process and development for the HCPB TBM manifold area elements, based on GTAW technique, are discussed. Manufacturing tests are performed using the EUROFER97 steel (RCC-MRx, AFCEN Code rules [ 1 ]). As part of the WCLL conceptual design development activities, an assembly and manufacturing welding sequence is detailed, focusing on design changes and the impact on the selection of manufacturing technologies. The article will present as well the future fabrication developments and the associated design proposals envisaged on TBM.

Long-term corrosion behavior of EUROFER RAFM steel in static liquid Pb-16Li at 550 °C

Liquid Pb-16Li eutectic alloy is the tritium breeding material in several breeding blanket concepts for European DEMO reactor, which are the Helium-Cooled Lithium-Lead (HCLL), the Water-Cooled Lithium-Lead (WCLL) and the Dual-Coolant Lithium-Lead (DCLL). RAFM (Reduced-Activation Ferritic/Martensitic) steels are considered as structural materials for breeding blankets in fusion reactors and EUROFER steel (9Cr-1.1W-0.2 V) is the European current choice for DEMO. Moreover, the WCLL European Test Blanket Module (TBM) to be installed in ITER will be manufactured in EUROFER. Here, the chemical compatibility of EUROFER in Pb-16Li is a major issue due to the dissolution of the material at high temperature in the liquid metal. In this work the corrosion of EUROFER in static Pb-16Li melt was investigated at 550 °C up to 8000 h of exposure to approximate low-flow WCLL blanket conditions. SEM-EDX analysis were performed on corroded samples to assess morphology and compositional changes occurring during the exposure. The corrosion was numerically evaluated by weight loss on samples exposed to various exposure times (2000, 4000 and 8000 h) and the corrosion rate for one year of exposure (8760 h) was extrapolated. The analysis by SEM-EDX showed that EUROFER mostly dissolved uniformly with a depletion of Cr and an enrichment of W at the near interface. The analysis of the weight loss provided an extrapolation of the corrosion rate of 18 μm/yr at 550 °C.

European TBM programme: First elements of RoX and technical performance assessment for DEMO breeding blankets

Abstract The strategy for the development of the European DEMO Breeding Blanket (BB) and the TBM (Test Blanket Module) Programme have been recently realigned with the objective to mitigate the risks associated with the development of tritium breeding and power extraction technologies in a DEMO Fusion Plant. In such a context, a clear definition of the key technical requirements of the DEMO Breeding Blanket (BB) is essential not only to drive a consistent BB design but also to allow the TBM Programme to provide its contribution to the preliminary demonstration of compliance with those requirements. Such contribution is substantiated through all phases of the TBM Programme life cycle, from the TBS (Test Blanket System) conceptual design up to its installation in ITER, operation and final dismantling. After a recall of the status of development of the European BB DEMO and TBM Programme, the paper describes the strategy and the status of the activities for the preliminary demonstration of compliance with a selected group of DEMO BB requirements. Through the elements given in this paper it is highlighted the importance to have a continuous and strict connection between the DEMO BB development and the execution of the TBM Programme.

Preliminary RAMI assessment for HCPB and HCLL ITER test blanket module ancillary systems

A RAMI (Reliability, Availability, Maintainability and Inspectability) assessment performed on the ITER Test Blanket Module (TBM) ancillary systems is presented. The assessment is aimed at evaluating design that may jeopardize the achievement of the overall 75% availability requirements for the TBM plant breakdown structure. The ancillary systems of the European Test Blanket Modules to be operated in ITER that have been analyzed in this assessment are the helium cooling systems (HCSs) of both the Helium Cooled Lithium Lead (HCLL) and Helium Cooled Pebble Bed (HCPB), the lithium-lead loop of the HCLL TBM and the Tritium Extraction System (TES) of the HCPB TBM. RAMI performance was assessed by means of reliability block diagrams (RBDs) over a foreseen mission of 20 years, with operating cycles reflecting ITER schedule. In particular, a set of events leading to unavailability of the systems was initially defined by means of a failure mode and effect analysis. Then RBDs were implemented according to reliability-wise integration of the components included in such systems. Different RBDs were defined depending on the component being judged to impact normal operation or start-up operation mode. Systems analysis was performed exploiting a modular approach in order to elicit the relative contribution to overall availability so to support possible design improvement by highlighting critical sub-systems. In particular the unavailability of cooling, tritium extraction and trapping functions were assessed separately for HCLL and HCPB concepts. Finally the integrated impact of all ancillary systems on overall tokamak operational availability was estimated and the assessed systems, considering the current preliminary level of design development, appear able to meet the expected performance.

Conceptual design of the enhanced coolant purification systems for the European HCLL and HCPB test blanket modules

The Coolant Purification Systems (CPSs) is one of the most relevant ancillary systems of European Helium Cooled Lead Lithium (HCLL) and Helium Cooled Pebble Bed (HCPB) Test Blanket Modules (TBMs) which are currently in the preliminary design phase in view of their installation and operation in ITER.The CPS implements mainly two functions: the extraction and concentration of the tritium permeated from the TBM modules into the primary cooling circuit and the chemistry control of helium primary coolant.During the HCLL and HCPB-TBSs (Test Blanket Systems) Conceptual Design Review (CDR) in 2015 it was recognized the need of reducing the tritium permeation into the Port Cell #16 of ITER. To achieve this and, then, to lower the tritium partial pressure in the Helium Cooling Systems in normal operation, the helium flow-rate treated by each CPS has been increased of almost one order of magnitude.In 2017, to satisfy the CDR outcomes and the new design requirements requested by Fusion for Energy (F4E, the European Domestic Agency for ITER), ENEA performed a preliminary design of the “enhanced” CPSs.This paper presents the current design of the “enhanced” CPSs, focusing on design requirements, assumptions, selection of technologies and preliminary components sizing.

Experimental Assessment of a Flat Sandwich-Like Self-Powered Detector for Nuclear Measurements in ITER Test Blanket Modules

Neutron and gamma flux measurements in designated positions in the test blanket modules (TBMs) of ITER will be important tasks during its campaigns. Investigations on self-powered detectors (SPDs), a class of reactor flux monitors are undertaken in the framework of an ongoing project on development of nuclear instrumentation for European ITER TBMs. This paper reports the findings of experiments performed with an SPD in flat sandwich-like geometry. A detector with vanadium emitter is chosen for preliminary studies. Its irradiation in a thermal neutron field gives a proof of the principle of flat SPDs. It is further irradiated in the mixed neutron–gamma field of a 14-MeV neutron generator and a bremsstrahlung photon field. The test SPD signals are proportional to the incident fluxes, deeming it suitable for flux monitoring. Whereas both neutrons and gammas can be detected with appropriate optimization of geometries, materials, and sizes of the components, the present design is more sensitive to gammas than fast neutrons. Based on the measured sensitivities of the SPD, its response under TBM conditions is predicted.

Computational study of a chromium self-powered detector for neutron flux monitoring in the test blanket module of ITER

Self-powered detectors (SPD) are online detectors for the monitoring of neutron and photon fluxes in nuclear reactors. The simplicity of design and operation of an SPD, its compactness, low-cost, reliability, and durability make it a candidate for active detectors in the European test blanket modules (TBM) of ITER. In this paper, we analyze a chromium emitter based SPD, which is proposed as a fast neutron detector for integration in the European TBMs. We describe a computational technique with Monte-Carlo modeling to predict the response of an SPD in the ITER environment and present the results of a demonstrative study. At first, the neutron and photon flux intensities and their energy spectra are calculated in the TBM through simulations in an MCNP model of ITER. Then, a multi-step multi-particle transport approach is adapted for calculation of the SPD sensitivity to the neutron and the photon spectra. Starting with the information on the composition of the SPD layers, main processes involved in the creation of the detector current are evaluated, corresponding to each of which a partial sensitivity is determined. Finally, a time-profile of the signal and its various components are constructed for one plasma pulse of ITER, with typical duration and fusion power. The range of electric currents achievable with SPDs is obtained, its adequacy for flux monitoring in TBMs is discussed, and finally, some design improvements are suggested.

Update in the nuclear responses of the European TBMs for ITER during operation and shutdown

The depiction of the nuclear responses of the ITER European Test Blanket Modules (TBMs), Helium Cooled Lithium Lead (HCLL) and Helium Cooled Pebbles Bed (HCPB) is presented in this work. Following important components update, and important methodological advances, the nuclear heat and the tritium production have been revisited, giving new estimations 10% higher than the previous evaluation for nuclear heat in both TBMs and to 15% higher for HCPB T production. This has an impact on the thermo-mechanical design of the TBM and the tritium handling. In addition, the Shutdown Dose Rates in the respective port interspace have been characterized in local approach. It shows a performance that could imply compatibility with planned in-situ maintenance activities when analysed in global approach, an improvement with respect to previous evaluations.

The European ITER Test Blanket Modules: Fabrication R&D progress for HCLL and HCPB

Two concepts have been chosen to be tested in ITER under the form of Test Blanket Modules (TBMs): the Helium-Cooled Lithium-Lead (HCLL) and the Helium-Cooled Pebble-Bed (HCPB). Both European TBMs designs share similar steel box structure which is constituted by a box, made of two Side Caps (SCs) and a First Wall (FW), stiffened by horizontal and vertical Stiffening Plates (SP) and closed on its back by several back plates (BPs). All structure subcomponents are internally cooled by Helium circulating in meandering squared section channels. This paper describes manufacturing technologies developed and implemented to assembly the SPs into the box.It presents the preliminary manufacturing procedure developed and applied for the assembly of the SPs into the box by Tungsten Inert Gas (TIG). Several mock-ups have been manufactured from laboratory to feasibility mock-ups (scale 1:1) on which non-destructive and destructive tests have been carried-out to identify the preliminary manufacturing procedure. Due to TBM specificities (namely complex welding trajectories, heavy and big components, plates with channels, space constraints, …) a specific welding facility including a custom welding torch and an automated bench has been achieved and is also described in the paper.We detail the adopted manufacturing strategies, as the optimization of welding sequence to minimize distortions and the customization of welding parameters, to compensate machining tolerances and welding gaps. Results such as welded joints quality and microstructure, internal cooling channel deformation and structure distortions are reported. These developments have been performed following a standardized procedure complying with professional codes and standards (RCC-MRx).

Analysis of test blanket module attachments under electromagnetic loads by using a dynamic amplification factor envelope

European test blanket modules (TBMs) are attached to radiation shields through mechanical joints. The structural integrity of the joints must be ensured under non-periodic dynamic loading caused by accidental electromagnetic events. This requires the transient dynamic analysis of the TBM-Joint-Shield under multiple loading scenarios. As temperatures and dynamic characteristics of the TBM-Joint-Shield system are updated under the course of the ITER project, the design process of the joints can become very time-consuming. This work presents a methodology based on a response spectrum to simplify the design process of the joint under electromagnetic loads. The mechanical system TBM-Joint-Shield was modeled as a single degree of freedom lumped spring-mass-dashpot system. The dynamic amplification was determined for a wide range of frequencies using the Finite Element Method. Eventually, a design envelope for each loading category was obtained, which can be readily used for determining the dynamic amplification factor of any potential new design of the TBM + Joint + Shield system. Results showed that helium cooled lithium lead (HCLL) undergoes higher loads than helium cooled pebble bed (HCPB), resulting in a dynamic amplification factor which varies from 1.25 to 2.0 for a damping of 4%.

Non-linear assessment of critical failure modes in the first wall of the European TBM

The assessment of the different components of the Helium Cooled Pebble Bed (HCPB) and Helium Cooled Lithium Lead (HCLL) Test Blanket Modules (TBMs) based on elastic FE analyses has been performed considering main failure modes and respective structural design criteria. This study has been carried out for the operation at full power and for the so-called Approach 2 cooling that is set up at reduced heat extraction capability for compliance with temperature targets to be relevant to DEMOnstration power plant blanket system. Consequently this situation is the worst condition in terms of cooling performance as a result it reveals that many locations in the First Wall (FW) become critical in the sense that certain criteria are not fulfilled, particularly those for immediate plastic flow localization, progressive deformation (ratcheting) and creep-fatigue interaction. Of course one can better reach code compliance by improving the cooling performance (so called Approach 1) but it will degrade the DEMO relevancy.The critical failure modes are re-assessed performing non-linear analysis using an elasto-viscoplatic model, particularly developed for RAFM steels such as the structural material EUROFER97 by Aktaa & Schmitt, and considering the appropriate design criteria of the inelastic route of RCC-MRx and SDC-IC. Despite the conservatism in these criteria, the results show all-clear signal with respect to immediate plastic flow localization and at least much smaller breaches of the limits with respect to the other failure modes. Reporting details and results of the non-linear failure assessments demonstrates their potential and capability in supporting specific and efficient improvements to the design of highly loaded components. The conservatism in the considered design criteria and ways for its reduction are discussed in addition.

Studies on Flat Sandwich-type Self-Powered Detectors for Flux Measurements in ITER Test Blanket Modules

Neutron and gamma flux measurements in designated positions in the test blanket modules (TBM) of ITER will be important tasks during ITER’s campaigns. As part of the ongoing task on development of nuclear instrumentation for application in European ITER TBMs, experimental investigations on self-powered detectors (SPD) are undertaken. This paper reports the findings of neutron and photon irradiation tests performed with a test SPD in flat sandwich-like geometry. Whereas both neutrons and gammas can be detected with appropriate optimization of geometries, materials and sizes of the components, the present sandwich-like design is more sensitive to gammas than 14 MeV neutrons. Range of SPD current signals achievable under TBM conditions are predicted based on the SPD sensitivities measured in this work.

Experimental Investigation of ZrCo Getter Beds as Candidate Process for the Tritium Extraction Systems of the European Test Blanket Modules

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 response of the material, controlled by temperature and H2 concentration, is the main limiting factor for the component performance.

Characterisation and radiolysis of modified lithium orthosilicate pebbles with noble metal impurities

Modified lithium orthosilicate (Li4SiO4) pebbles with additions of titanium dioxide (TiO2) are suggested as an alternative tritium breeding ceramic for the European solid breeder test blanket module. The noble metals – platinum (Pt), gold (Au) and rhodium (Rh), can be introduced into the modified Li4SiO4 pebbles during the melt-based process, due to the corrosion of Pt-Rh and Pt-Au alloy crucible components. In this study, the surface microstructure, chemical and phase composition of the modified Li4SiO4 pebbles with different contents of the noble metals was analysed. The influence of the noble metals on the radiolysis was evaluated after irradiation with accelerated electrons (E=5MeV), up to 12MGy absorbed dose at 300–345K in a dry argon atmosphere. Using electron spin resonance (ESR) spectroscopy, it was determined that the noble metals (up to 300ppm) do not significantly influence the formation and accumulation of radiation-induced defects (RD) in the modified Li4SiO4 pebbles.

PERMCAT experiments with tritium at high helium flow rates relevant for the tritium extraction systems using the CAPER facility at TLK

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.