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.

The ITER Isotope Separation System (ISS) and Water Detritiation System (WDS) should be integrated in order to reduce potential chronic tritium emissions from the ISS. This is achieved by routing the top (protium) product from the ISS to a feed point near the bottom end of the WDS Liquid Phase Catalytic Exchange (LPCE) column. This provides an additional barrier against ISS emissions and should mitigate the memory effects due to process parameter fluctuations in the ISS.To support the research activities needed to characterize the performances of various components for WDS and ISS processes under various working conditions and configurations as needed for ITER design, an experimental facility called TRENTA representative of the ITER WDS and ISS protium separation column, has been commissioned and is in operation at TLK.The experimental program on TRENTA facility is conducted to provide the necessary design data related to the relevant ITER operating modes. The operation availability and performances of ISS-WDS have impact on ITER fuel cycle subsystems with consequences on the design integration. The preliminary experimental data on TRENTA facility are presented.

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 of focussed 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.

ITER is a large vacuum facility which comprises many service, diagnostic and monitoring vacuum sub-systems as well as three large cryogenic pumping systems for evacuation and maintenance of the requested pressure levels. The presence of hydrogen (including the radioactive isotope tritium) and exclusion of other gases defines the parameters of fusion vacuum systems. This paper will focus on the areas of the ITER vacuum systems which require customized developments and cannot rely on commercial solutions. The complex pumps have been tailored for the very specific applications and requirements at ITER, characterized by high magnetic and radiation fields, excellent availability and maintainability and, especially, the need to be tritium-compatible. An outline of the development path which was needed to come up with a sound design for the ITER cryopumps is given and the status of the programme in view of the imminent manufacturing phase is described in this paper.

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 set(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

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 nucleaires 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. (c) 2008 Elsevier B.V. All rights reserved.

The vacuum systems for ITER are characterized by the requirements for tritium compatibility, tolerance of high magnetic and radiation fields and remote maintainability. In addition, although the vacuum levels are relatively modest, high pumping speeds are needed to achieve the high gas throughputs required. The design solutions adopted, the status of the development programme and the issues still to be addressed before commitment of the designs to fabrication are described in the paper.

Many material selections for fusion Fuel Cycle systems are determined by the properties of tritium, including its behaviour as a hydrogen isotope, and its decay product, 3He. Within the EU R&D program, the following issues related to tritium service have been addressed. The mechanical integrity and longevity of the sorbent/bonding agent/substrate system used for cryosorption pumping have been extensively tested under tritium exposure. Extended testing of palladium/silver membranes used for separation of elemental hydrogens from impurities has been carried out to confirm longevity in tritium service. For all high temperature (∼150°C) components, tritium permeation through primary containments must be confined by outer (low temperature) jackets, and designs have been developed to achieve this. For wetproof catalysts and solid polymer electrolysers used for water detritiation, tests are in progress to determine the operating life. Testing of the ITER reference tritium storage getter material is under way.

The paper describes the key elements of the EU fuel cycle development for ITER. The programme is now closely focused on the systems allocated to the EU for in-kind supply. These include the isotope separation and water detritiation systems (ISS and WDS) and the cryopumps for torus exhaust, neutral beam injectors (NBI) and cryostat pumping. For all of these systems, which must be available for (or even before) first plasma operation, prototypes at ITER-relevant scale are under design and construction and the concepts, critical issues, experimental programmes and status of these development projects are outlined in the paper.

The ITER Isotope Separation System (ISS) and Water Detritiation System (WDS) will be integrated in order to reduce potential chronic tritium releases from the ISS by routing the top (protium) product from the ISS into the Liquid Phase Catalytic Exchange (LPCE) column of WDS. This provides an additional barrier against ISS tritium releases and should mitigate the memory effects due to process parameter fluctuations in the ISS.To support the research activities needed to characterize the performances of various components for WDS and ISS processes in various working conditions and configurations as needed for ITER design, an experimental facility called TRENTA and representative of the ITER WDS and ISS protium separation column has been commissioned at Tritium Laboratory Karlsruhe (TLK).The TRENTA facility consists of Combined Electrolysis Catalytic Exchange (CECE) process, with an LPCE column of 8 m, in combination with a cryogenic distillation (CD) process. The processes description and the status of commissioning of TRENTA facility is presented. (C) 2007 Elsevier B.V. All rights reserved.

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.

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.

Within the tritium plant of ITER a total inventory of about 2-3 kg will be necessary to operate the machine in the DT phase. During plasma operation, tritium will be distributed in the different sub-systems of the fuel cycle. A tool for tritium inventory evaluation within each sub-system of the fuel cycle is important with respect to both the process of licensing ITER and also for operation. It is very likely that measurements of total tritium inventories may not be possible for all sub-systems; however, tritium accounting may be achieved by modelling its hold-up within each sub-system and by validating these models in real-time against the monitored flows and tritium streams between the sub-systems. To get reliable results, an accurate dynamic modelling of the tritium content in each sub-system is necessary. A dynamic model (TRIMO) for tritium inventory calculation reflecting the design of each fuel cycle sub-systems was developed.The amount of tritium needed for ITER operation has a direct impact on the tritium inventories within the fuel cycle sub-systems. As ITER will function in pulses, the main characteristics that influence the rapid tritium recovery from the fuel cycle as necessary for refuelling are discussed.The confinement of tritium within the respective sub-systems of the fuel cycle is one of the most important safety objectives. The design of the deuterium/tritium fuel cycle of ITER includes a multiple barrier concept for the confinement of tritium. The buildings are equipped with a vent detritiation system and re-circulation type room atmosphere detritiation systems, required for tritium confinement barrier during possible tritium spillage events. Complementarily to the atmosphere detritiation systems, in ITER a water detritiation system for tritium recovery from various sources will also be operated.

The next generation thermonuclear fusion machine ITER comprises large cryogenic pumping systems to process very high gas throughputs. The paper starts from a summary of the requirements for the torus exhaust pumping system and the neutral beam injection system for plasma heating and discusses the design solutions derived to match them. All ITER force-cooled cryopumps are based on a modular set-up and incorporate similar design of charcoal-coated cryopanels cooled to 4.5K with supercritical helium. A systematic approach for how to design a tailor-made cryopump is presented. It is based on Monte Carlo simulation to calculate the overall capture coefficient being the primary design parameter. Experimental results measured with a model pump of representative scale are used to validate the design procedure. In the last part, the paper addresses the design issues to be solved for tritium-compatible forepumps with high pumping speeds.

Within the frame of the European Fusion Programme Forschungszentrum Karlsruhe is charged to develop the vacuum systems for ITER. These activities comprise the high vacuum systems for the torus exhaust gas, the neutral beam injectors and the cryostat, as well as the mechanical forepump trains for all of these systems. This paper describes the several systems and gives an overview about the underlying R&D work performed to provide a design basis for large vacuum pump systems. Most of the presented cryosorption results have been measured in the TIMO test bed, which houses an approximately 1:2 scale model of the torus cryosorption pump. In TIMO, almost all ITER relevant conditions can be replicated (like operational times, coolant flows, gas flows, temperatures, pressure), except of the pumping of tritium. The latter issue will be dealt with in a dedicated experiment during the 2003 trace tritium campaign in JET. It is shown that cryosorption vacuum pumping is an efficient and robust technique to be employed for ITER. Concepts for an appropriate forepump train for the high vacuum pumps are also discussed.

Currently, the strategy for determination of ITER in-vessel tritium inventory envisages that at predetermined intervals, tritiated gases in all systems of fuel cycle will be transferred to the storage and delivery system (SDS) and tritium quantities measured by in-bed calorimetry. The isotope separation system (ISS) is the system used to separate hydrogen isotopes at the quality required to be stored in SDS, and is one of the systems with highest tritium inventory within the fuel cycle. Therefore, during tritium inventory procedure, ISS has to be ‘milked down’ of tritium, mainly as DT molecular species. Based on the dynamic modelling code TRIMO of the tritium content in the main sub-systems of ITER Fuel Cycle, the procedure for tritium extraction from ISS is presented and numerical examples given to assess the necessary time of transferring the tritium from ISS to SDS, and the residual amount of tritium in ISS after different milking scenarios. Consequently a fuel handling strategy during tritium inventory assessment in the ISS and SDS is described, with the constraint of mobilizable tritium inventory minimisation.

The ITER in-vessel hydrogen inventory is highly dynamic and determined by a broad range of parameters. Moreover, it is of prime importance to allocate the inventory as accurately as possible. Two essential control issues have to be met. The basic safety approach is to limit the amount of hydrogen such that the resulting pressure in case of an explosion accident is compatible to the design criteria of the vacuum vessel. This requirement drives the operational philosophy and the design of the cryogenic pumps which are connected to the vessel. The administrative limitation of the tritium inventory is another essential requirement on reactor operation. This paper illustrates how hydrogen inventory control issues are reflected by the operational schemes of the vacuum pumping systems. Latest results include the release characteristics of hydrogens and tritium during normal regeneration of the cryopumps and under off-normal events. The different semi-permanent contributions to the overall tritium inventory are discussed. The forepumping systems reduce the in-vessel inventory and act as interface to the tritium plant. A typical example of an integral pump-out curve is discussed. It is shown that sufficient safety margin is included in the ITER vacuum pumping design.

The PERMCAT is a membrane reactor proposed for processing fusion reactor plasma exhaust gas: tritium removal is obtained by isotopic swamping operating in counter-current mode. In this work, a membrane reactor using a permeator tube of length about 500mm produced via diffusion welding of Pd-Ag thin foils is described. An appropriate mechanical design of the membrane module has been developed in order to avoid any significant compressive and bending stresses on the very long and thin wall permeator tube: two expanded bellows have been applied to the Pd-Ag tube, so that it has been pre-tensioned before operating. The elongation of the metal permeator under hydrogenation has been theoretically estimated and experimentally verified for properly designing the membrane reactor. (c) 2006 Elsevier B.V. All rights reserved.

A method to localise water leaks from the water cooling circuits into the ITER vacuum vessel has been demonstrated experimentally. The process involves adding a tracer material to the suspect cooling circuit and then looking for the tracer in the gases released after regenerating a torus cryopump at temperatures well above ambient. The detection technique used was gas chromatography-mass spectrometry and the method was demonstrated on the TIMO facility at the Forschungszentrum Karlsruhe, a half-scale model of the proposed ITER cryopumps.

Within the European research activities for the ITER fusion reactor, at the Forschungszentrum Karlsruhe (FZK) a test bed for the half-scaled ITER cryosorption pump was installed to study all the operational modes of interest. Following an extensive test program under ITER-like throughput conditions, additional studies were performed to assess the influence of liquid impurities on the sorption behaviour. After an extensive test program could already be successfully carried out the influences of liquid impurities on the pump performance should be studied. Tests with both water-like impurities and different tracer candidates were used during ITER-relevant operation cycles. Independent of the different tested impurities it could be proved that all requirements for the primary ITER torus vacuum pumps can be fulfilled with the actual vacuum pump concept. The total test program demonstrated that the adopted design concept meets all the ITER pumping requirements.