M. Glugla

Karlsruhe Institute of Technology, Carlsruhe, Baden-Württemberg, Germany

Are you M. Glugla?

Claim your profile

Publications (105)98.27 Total impact

  • [Show abstract] [Hide abstract]
    ABSTRACT: Consecutive absorption/desorption cycles of the ZrCo-H₂ system were studied to simulate the real International Thermonuclear Experimental Reactor (ITER) hydrogen getter system. A ZrCo getter was used in this paper instead of the depleted uranium (DU) getter material, which has been recently considered as the hydrogen getter in ITER. In a cyclic pressure-composition isotherm (PCI) measurement, the high-pressure Sievert apparatus seems impractical to describe the equilibrium state of the ZrCo-H₂ system in detail, especially for the desorption stage. This high-pressure PCI apparatus, however, shows cause and effect well, from the previous getter state to the following state in presenting hydriding/dehydriding performance. In case of the ZrCo-H₂ system or in case of the DU-H₂ system having multiple getter bed battery, a similar affection by previous getter status might be related and a similar aspect could be shown to consider further ITER design; for example, a need for control logic from PCI measurements using a high-pressure Sievert apparatus.
    No preview · Article · Jul 2015 · IEEE Transactions on Plasma Science
  • [Show abstract] [Hide abstract]
    ABSTRACT: Three kinds of Pt-catalyzed zeolite were tested as candidates for isotopic exchange of highly tritiated water (HTW), and CBV 100 CY (Na-Y, Si/Al∼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.
    No preview · Article · Apr 2015 · Fusion Science and Technology
  • [Show abstract] [Hide abstract]
    ABSTRACT: Highly tritiated water (HTW) may be generated at ITER by various processes and, due to the excessive radio toxicity, the self-radiolysis and the exceedingly corrosive property of HTW, a potential hazard is associated with its storage and process. Therefore, the capture and exchange method for HTW utilizing Molecular Sieve Beds (MSB) was investigated in view of adsorption capacity, isotopic exchange performance and process parameters. For the MSB, different types of zeolite were selected. All zeolite materials were additionally platinized. The following work comprised the selection of the most efficient zeolite candidate based on detailed parametric studies during the H2/D2O laboratory scale exchange experiments (∼25 g zeolite per bed) at the Tritium Laboratory Karlsruhe (TLK). For the zeolite, characterization analytical techniques such as Infrared Spectroscopy, Thermogravimetry and online mass spectrometry were implemented. Followed by further investigation of the selected zeolite catalyst under full technical operation, a MSB (∼22 kg zeolite) was processed with hydrogen flow rates up to 60 mol h-1 and deuterated water loads up to 1.6 kg in view of later ITER processing of arising HTW.
    No preview · Article · Apr 2015 · Fusion Science and Technology
  • [Show abstract] [Hide abstract]
    ABSTRACT: The CAPER facility of the Tritium Laboratory Karlsruhe has demonstrated the technology for the tokamak exhaust processing. CAPER has been significantly upgraded to pursue R&D towards highly tritiated water (HTW) handling and processing. The preliminary tests using a metal oxide reactor producing HTW afterward detritiated with PERMCAT were successful. In a later stage, a micro-channel catalytic reactor was installed in view of long term R&D 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 mu-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 platinized 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.
    No preview · Article · Mar 2015 · Fusion Science and Technology
  • [Show abstract] [Hide abstract]
    ABSTRACT: The main objective of ITER tritium Storage and Delivery System (SDS) is contracted to develop an optimal metal hydride bed that can be reveal the unprecedented fueling performance for the Tokamak. One function of the hydride bed is to keep safety requirements in terms of confinement of tritium. The hydride material for storing the deuterium and tritium fuelling gases is being made narrow with depleted uranium (DU) by its good performance. DU also has its own uncertainties, however, in applying it to realize the getter bed system having an all-round capability, especially in aspect of safety. This paper deals with from bed design target to the design variables in terms of comparison of risk-based multi-criteria using HAZOP (risk matrix) analysis. In analysis of the risks, important variables that denotes safety-effective, or cost-effective, or maintainability-effective, or manufacturability-effective are sometimes mutually interrelated with each other. As a conclusion the authors could recommend the way to concentrate and minimize the bed design variables with most meaningful risk-containing components that can be applied to increase the performance of hydride bed. It needs, however, that further study of comparison of risk analyses should be proceeded to complete the hydride bed design.
    No preview · Article · Oct 2014 · Fusion Engineering and Design
  • [Show abstract] [Hide abstract]
    ABSTRACT: The Storage and Delivery System (SDS) of the ITER Tritium Plant has to safely handle the fuel gases including tritium and deliver those gases to the Fuelling System (FS). Recently the ITER fuelling scenarios have been developed in more detail considering ramp-up, flat-top, and ramp-down. With this as input, an alternative analysis was performed for how SDS will support ITER inductive, hybrid, and non-inductive plasma operations. The fuelling rates from SDS to FS were evaluated. To supply gas to FS, SDS must draw gases from one or more sources. These sources could be SDS tanks, SDS hydride storage beds or the Isotope Separation System. Case studies were performed to evaluate the relative merits on various configurations. For inductive operations, it was found that tritium could be supplied with either 27 hydride beds and one tank or with 12 beds and four tanks. For deuterium supply the results were either 43 beds and one tank or 31 beds and four tanks. Also studied were options for distributing supporting gas inventories elsewhere in the Fuel Cycle or on larger hydride beds. Evaluation criteria included operability and safety.
    No preview · Article · Oct 2014 · Fusion Engineering and Design
  • [Show abstract] [Hide abstract]
    ABSTRACT: Consecutive absorption/desorption cycles of the ZrCo-H2 system were studied to simulate the real ITER hydrogen getter system. ZrCo getter was used in this study instead of the depleted uranium (DU) getter material which was recently considered as the hydrogen getter in ITER. In a cyclic PCI measurement the high-pressure Sievert apparatus seems impractical to describe the equilibrium state of the ZrCo-H2 system in detail, especially for the desorption stage. This high-pressure Pressure-Composition Isotherm (PCI) apparatus, however, shows a cause-and-effect well, from the previous getter state to the following state in presenting hydriding/dehydriding performance. In case of the ZrCo-H2 system or in case of the DU-H2 system, having multiple getter bed battery, a similar affection by the previous getter status might be related and a similar aspect could be shown to should consider further in ITER design, for example a need for control logic, from PCI measurements using a high-pressure Sievert apparatus.
    No preview · Article · Jan 2011
  • [Show abstract] [Hide abstract]
    ABSTRACT: A PERMCAT reactor is a catalytic membrane reactor that combines a Pd/Ag membrane and a catalyst bed. It has been developed to ensure very high tritium recovery from the unspent fuel of fusion machines using deuterium tritium mixtures. The PERMCAT process takes advantage of simultaneously unlocking chemically bound tritium via heterogeneously catalysed isotope exchange reactions and removing tritium via its selective permeation through the membrane. The PERMCAT reactor operated in the counter-current isotope swamping mode allows a very low tritium activity at the outlet of the component to be maintained.Two main issues have been solved to achieve efficient and reliable PERMCAT operation. Firstly, the mechanical design has to cope with the elongation and deformation of the membrane resulting from thermal expansion and lattice parameter increase under operation with hydrogen. Secondly, the catalyst material has to be chosen in order to promote isotope exchange reactions while minimising the numerous side reactions that occur especially when the mixture contains carbon oxides. This paper presents a general overview of the R&D performed at the Tritium Laboratory Karlsruhe for PERMCAT technology. Technical solutions to solve both issues together with relevant experimental results including processing tests with tritium are discussed.
    No preview · Article · Oct 2010 · Catalysis Today
  • [Show abstract] [Hide abstract]
    ABSTRACT: The ITER Nuclear Buildings include the Tokamak, Tritium and Diagnostic Buildings (Tokamak Complex) and the Hot Cell and Low Level Radioactive Waste Buildings (Hot Cell Complex). The Tritium Confinement strategy of the Nuclear Buildings comprises key features of the atmosphere and vent detritiation systems and the heating, ventilation and air conditioning systems. The designs developed during the ITER EDA (engineering design activities) for these systems need to be adapted to the specific conditions of the Cadarache site and modified to conform with the regulatory requirements applicable to installations nucléaires de base (INB) – basic nuclear installations – in France. The highest priority for such adaptation has been identified as the Tritium Confinement of the Tokamak Complex and the progress in development of a robust, coherent design concept compliant with French practice is described in the paper.
    No preview · Article · Dec 2008 · Fusion Engineering and Design
  • [Show abstract] [Hide abstract]
    ABSTRACT: The storage of hydrogen isotopes as metal hydride is the technique chosen for the ITER Tritium Plant Storage and Delivery System (SDS). A prototype storage bed of a full-scale has been designed, manufactured and intensively tested at the Tritium Laboratory, addressing main performance parameters specified for the ITER application. The main requirements for the hydrogen storage bed are a strict physical limitation of the tritium storage capacity (currently 70 g T2), a high supply flow rate of hydrogen isotopes, in-situ calorimetry capabilities with an accuracy of 1 g and a fully tritium compatible design. The pressure composition isotherm of the ZrCo hydrogen system, as a reference material for ITER, is characterised by significant slope. As a result technical implementation of the ZrCo hydride bed in the SDS system requires further considerations. The paper presents the experience from the operation of ZrCo getter bed including loading/deloading operation, calorimetric loop performance, and active gas cooling of the bed for fast absorption operation. The implications of hydride material characteristics on the SDS system configuration and design are discussed.
    No preview · Article · Jul 2008 · Fusion Science and Technology
  • U. Besserer · L. Dörr · M. Glugla
    [Show abstract] [Hide abstract]
    ABSTRACT: This paper describes the tritium confinement concept and the tritium retention systems at TLK. A description of the AMOR facility for the regeneration of the HTO loaded molecular sieve beds and the operational experience gained from the regeneration of molecular sieve beds (up to 20 times each) is also presented. Finally tritium releases over this period to the environment will also be given.
    No preview · Article · Jul 2008 · Fusion Science and Technology
  • [Show abstract] [Hide abstract]
    ABSTRACT: Tritium as one of the two fuel components for fusion power plays a special role in any fusion device. Due to its volatile character, radioactivity and easy incorporation as HTO it needs to be controlled with special care and due to its scarcity on earth it has to be produced in-situ in future fusion power plants. The paper discusses the present tritium R&D activities in fusion ongoing in the EU and presents the various processes/techniques envisaged for controlling tritium in future fusion reactors focusing mainly on the issues of breeding blankets and the fuel cycle in DEMO.
    No preview · Article · Jul 2008 · Fusion Science and Technology
  • [Show abstract] [Hide abstract]
    ABSTRACT: One of the key activities on ITER during 2007 is a Design Review covering selected high priority areas of the project in which a significant number of features of the design with the potential to compromise the achievement of some objectives of ITER have been identified. These issues are being addressed by a number offocussed working groups to develop solutions for these issues which will enhance operating margins, reliability and availability, and ensure compliance with the French licensing framework. One of the working groups has been set up to investigate tritium-related issues. The principal design features which are being addressed by this group and the proposed resolutions of these issues are described in the paper.
    No preview · Article · Jul 2008 · Fusion Science and Technology
  • [Show abstract] [Hide abstract]
    ABSTRACT: Tracking of tritium inventories on ITER will be essential to ensure that the safety limits established for the mobilizable tritium inventory in the vacuum vessel are not violated. Tritium will be delivered to the ITER site from outside suppliers. Staring with the tritium imports the value of tritium inventory at ITER site will be known with a certain error that will propagate in time. During plasma operation, shot by shot measurements of the tritium delivered to the Torus and recovered will allow the amount of tritium trapped in the Torus to be computed at the end of the day. A case study for different measuring techniques and several measuring points for the tritium recovered from Torus have been done. An alternative method is to measure overnight the variation in the inventory of the storage and delivery system and the associated error when this method will be employed are presented. In order to reduce the errors on the tritium trapped in-vessel, at certain time intervals a method of global tritium inventory will be performed. The method envisages the transfer of all the mobilizable tritium from the plant and measurement of this inventory in the self-assay beds from the storage and delivery system. Evaluation of the most important sources of error for the tritium trapped in-vessel and means of minimization are eventually presented.
    No preview · Article · Jul 2008 · Fusion Science and Technology
  • [Show abstract] [Hide abstract]
    ABSTRACT: 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.
    No preview · Article · Jul 2008 · Fusion Science and Technology
  • [Show abstract] [Hide abstract]
    ABSTRACT: The design of tritium processing loop for KATRIN tritium source supported by high purity of tritium has been demonstrated. The demonstration has showed tritium beta spectrum using a high energy resolution electrostatic spectrometer with adiabatic magnetic collimation. The demonstration proved that the systematic uncertainties in KATRIN can cause the use of a windowless gaseous tritium source (WGTS) as the main beta source. The tritium processing system (TPS) involves an inner loop providing the circulation of tritium through WGTS and control of isotope composition of the circulating gas. The demonstration compress and purify the pumped tritium by TPS and returned back to the source. The processing capability of the system in the demonstration kept on the level of 40 gram of tritium per day. The demonstration proved that the system can fulfil the requirements of KATRIN project with stable circulation of pure tritium.
    No preview · Article · Jul 2008 · Fusion Science and Technology
  • [Show abstract] [Hide abstract]
    ABSTRACT: 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.
    No preview · Article · Jul 2008 · Fusion Science and Technology
  • [Show abstract] [Hide abstract]
    ABSTRACT: During plasma operation of ITER in the DT phase, tritium will be distributed in the different subsystems of the fuel cycle; tritium inventories within the systems are not constant, but vary as the gas moves through these systems during the burn and dwell periods. To evaluate the tritium content in each sub-system of the fuel cycle of ITER, a dynamic model for tritium inventory calculation was developed. The code reflects the design of each system in various degrees of detail; both the physical processes characteristics and in some cases the associated control systems are modeled. The amount of tritium needed for ITER operation has a direct impact on the tritium inventories within the fuel cycle subsystems. As ITER will function in pulses, the main characteristics that influence both the maximum value of tritium inventories in the systems and the rapid tritium recovery from the fuel cycle as necessary for refueling are discussed. Eventually the inventories in the Isotope Separation System (as the system with the highest tritium inventory) for short and long pulses and their dependence on the packing molar inventory are presented.
    No preview · Article · Oct 2007
  • D. Demange · S. Welte · M. Glugla
    [Show abstract] [Hide abstract]
    ABSTRACT: The PERMCAT process chosen for the final clean-up stage of the Tokamak Exhaust Processing system of the ITER tritium plant combines in a single component a catalytic reactor and a permeator using Pd/Ag membranes. This study covers the mechanical behaviour of a Pd/Ag membrane under different operating conditions. The consequences of hydrogen uptake by the membrane during nominal operation but also during off-normal events are presented. Depending on the operating conditions, expansions around 2% and significant deformations are observed. Different mechanical designs of PERMCAT reactors are then discussed. The first generation comprises finger-type membranes and two new mechanical designs use either additional edge welded bellows or a special corrugated Pd/Ag membrane. These upgraded designs improve the robustness and simplify the geometry of the component. The experimental validation of these new units has been carried out based on the measurements of the processing capabilities with regards to isotopic exchanges between H2O and D2. For the first time experimental results obtained on different PERMCAT reactors are compared.
    No preview · Article · Oct 2007 · Fusion Engineering and Design
  • [Show abstract] [Hide abstract]
    ABSTRACT: One of the main concerns related to licensing of ITER is the amount of potentially tritium release into the environment and the qualification of the barriers against tritium release. The final barrier of tritium release from fuel cycle is the Water Detritiation System (WDS) which will be operated in combination with the Isotope Separation System (ISS). To investigate the performances of various components of these systems, an experimental facility based on Combined Electrolysis Catalytic Exchange (CECE) process with a Cryogenic Distillation (CD) process was built at Tritium Laboratory Karlsruhe. The investigations are focused on two main issues: to quantify the separation performances of deuterium and tritium within the Liquid Phase Catalytic Exchange (LPCE) and CD processes in steady state and in dynamic mode of operation and to develop an integrated control system to be used in ITER ISS, in order to minimize the tritium inventory and to reduce at maximum extent the tritium releases. At TLK the two systems, CECE and CD have been commissioned and the experimental program and preliminary functionality tests of the main components are presented.
    No preview · Article · Oct 2007 · Fusion Science and Technology