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.

Hydrogen isotopes from the Water Detritiation System (WDS) developed at the Tritium Laboratory Karlsruhe (TLK) are fed to a cryogenic distillation (CD) column. In order to purify this gas a permeator of 1m2 membrane area was found to be necessary. Technical requirements, regulations of the European Union and the TLK safety philosophy demanded that the component had to be designed and manufactured in accordance to the “Technical Terms of Delivery and Acceptance” for tritium primary systems in the Tritium Laboratory Karlsruhe and in conformity to the European Pressure Equipment Directive (PED). Since there was no response to the call for tender from industrial partners, the permeator was manufactured at the main workshop of the Forschungszentrum Karlsruhe (FZK). Acceptance tests and commissioning showed that the component performance is better than design requirements. This paper describes the construction and commissioning of the permeator and gives the results of the commissioning tests.

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.

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.

The research for the performance improvement of the liquid phase chemical exchange (LPCE) column has been carried out at Nagoya University with the collaboration of National Institute for Fusion Science (NIFS) and Tritium Laboratory Karlsruhe (TLK). Kogel catalysts and Dixon gauze rings were mixed at a certain ratio and packed in the column homogeneously. A stage-wise model, named “Channeling stage model”, was also developed to predict separative performances of the column. In order to design a LPCE column to be tested at the TLK experimental facility, we examined the effect of the catalyst packed ratio and the gas–liquid ratio on the separative performance of the column. The optimal value of the catalyst packed ratio was obtained to 30%. The specifications of the column and the operating conditions of the equipment were reported.

Abstract The design of a tritium processing loop for KATRIN tritium source and results of test set-up operation are presented. The constant source intensity is supported by high purity (> 95% of tritium) of tritium circulated through the source with 4.8·10-4 g/s rate. The possibility of source parameters stabilization was verified with test facility “TILO” which was built for this purpose. The long term (30 days) deuterium circulation with flow rate 0.2 Pa m3/s shows that using of standard high accuracy pressure sensors from MKS and NI control electronics allows to stabilize the flow rate in the system with 0.1% accuracy. The reliability of process equipment was tested as well during about 3000 h of TILO operation.

The neutron capture cross section of 14C is of relevance for several nucleosynthesis scenarios such as inhomogeneous Big Bang models, neutron induced CNO cycles, and neutrino driven wind models for the r process. The 14C(n,g) reaction is also important for the validation of the Coulomb dissociation method, where the (n,g) cross section can be indirectly obtained via the time-reversed process. So far, the example of 14C is the only case with neutrons where both, direct measurement and indirect Coulomb dissociation, have been applied. Unfortunately, the interpretation is obscured by discrepancies between several experiments and theory. Therefore, we report on new direct measurements of the 14C(n,g) reaction with neutron energies ranging from 20 to 800 keV.

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.

These days more and more modern electrolysis cells are operated with new solid polymer membranes. These membranes prevailing DuPont's Nafion (R) are not only used for electrolysis but as well for the wide spectrum of fuel cells due to their good mechanical and chemical stability and the high proton conductivity. For that reason it is intended to use these solid polymer membranes for the electrolyzer units in the ITER Water Detritiation System (WDS). The influence of Tritium during water electrolysis to the membrane material is still not sufficiently investigated. Therefore long-term experiments ofsolidpolymer membranes were performed at Tritium Laboratory Karlsruhe (TLK). The chemical degradation and durability behavior of the used Nafion (R) 117 membrane are investigated in details under tritiated water conditions. For comparison a second cell was operated with demineralized water for the same period.In addition to the experimental rigs with single Nafion (R) membranes, two industrial electrolyzer units equipped with Nafion (R) membranes were operated during different tritium experiments at TLK. Before operation they had been modified to be compatible for tritium operation. After long operation period no degradation in the performance of the electrolyzers is observable.

ITER is the first fusion machine fully designed for operation with equimolar deuterium-tritium mixtures. The tokamak vessel will be fuelled through gas puffing and pellet injection, and the Neutral Beam heating system will introduce deuterium into the machine. Employing deuterium and tritium as fusion fuel will cause alpha heating of the plasma and will eventually provide energy. Due to the small burn-up fraction in the vacuum vessel a closed deuterium-tritium loop is required, along with all the auxiliary systems necessary for the safe handling of tritium.The ITER inner fuel cycle systems are designed to process considerable and unprecedented deuterium-tritium flow rates with high flexibility and reliability. High decontamination factors for effluent and release streams and low tritium inventories in all systems are needed to minimize chronic and accidental emissions. A multiple barrier concept assures the confinement of tritium within its respective processing components; atmosphere and vent detritiation systems are essential elements in this concept.Not only the interfaces between the primary fuel cycle systems - being procured through different Participant Teams - but also those to confinement systems such as Atmosphere Detritiation or those to fuelling and pumping - again procured through different Participant Teams - and interfaces to buildings are calling for definition and for detailed analysis to assure proper design integration. Considering the complexity of the ITER Tritium Plant configuration management and interface control will be a challenging task. (c) 2007 Elsevier B.V. All rights reserved.

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.

Approximately 1g of the 150g collected at the JET was send to the Tritium Laboratory Karlsruhe (TLK) for physico-chemical characterisation. The flakes were extensively analysed using several experimental tools, such as calorimetry, X-ray-diffraction, thermal desorption, helium pycnometry, scanning electron microscopy, etc. The specific surface area measurements according to the Brunauer–Emmett–Teller (BET) method, showed a value of about (4.7±0.3)m2g−1 while their average tritium activity was estimated to be (1.17±0.1)TBqg−1 (i.e. 3–3.3mg of tritium per gram of flakes) and their average real density (1.69±0.02)gcm−3. Based on these results we estimate that at the end of the remote tile exchange (RTE) at JET the 3.0g of tritium which did not return to the AGHS after the DTE1 remained in the vessel mainly in form of flakes, i.e. approximately (950±80)g of flakes are still inside the machine.

The confinement of tritium within its respective processing systems is clearly one of the most important safety objectives for fusion reactors. Consequently, the functional safety of all the subsystems of the inner deuterium/tritium fuel cycle of ITER need to be carefully analyzed and the results implemented into the design of the Tritium Plant as a whole. The basic principles for the management of functional safety are laid out in the international standard IEC 61508.The tritium bearing systems need to be particularly protected against over-pressure and over-temperature, even at levels significantly below values at which sensors or components would loose their mechanical integrity. The design shall take into account the necessity to validate and test the protection measures, noting the contamination of the equipment with tritium and the restricted access due to multiple barriers.Any protection measure has its characteristic but limited credibility. Software based safeguarding, for example, is not fully satisfactory in certain cases. In spite of the developments in software and the increasing possibilities to assure its quality, hard-wired signal processing is still considered to be necessary within the ITER Tritium Plant. (c) 2005 Elsevier B.V. All rights reserved.

Flakes and dust are generated by co-deposition in all experimental fusion machines having a carbon first wall. At JET a very large fraction of tritium is retained in form of flakes in the vicinity of the water cooled louvres adjacent to the inner divertor and in the sub-divertor zones of the machine. The BET specific surface area performed on flakes collected after a DD campaign confirm a low value of at about 7 m 2 g - 1 . The average tritium activity of flakes after DTE1 was estimated to be (1.17 ′ 0.1) TBq g - 1 . At the end of RTE approximately 3.0 g tritium was not accounted for and must remain in the vessel as flakes. If they have the same content of tritium as the flakes removed up to now then a further (950 ′ 80 g)of flakes must still be present in the vessel.

A Task Force Fuel Cycle has been established at the Forschungszentrum Karlsruhe, aiming to detail the design of the strongly interlinked deuterium/tritium systems of ITER. The ultimate result of the work shall be a complete set of drawings and descriptive documentation to allow manufacture and assembly of the fuel cycle systems by competent industrial partners. However, the responsibility for the performance of the tritium processing systems shall remain with the designers and shall not be passed to the manufacturers. Above and beyond the efforts on R&D and design of the analytical system, the storage and delivery system and the tokamak exhaust processing system of ITER the potential trades off between the isotope separation system and the water detritiation system of ITER are currently under experimental investigation at the Tritium Laboratory Karlsruhe (TLK). One of the aims is to reduce the effluents of ITER along with the ALARA (as low as reasonable achievable) process. The activities of the Task Force Fuel Cycle shall also cover the procurement packages for the cryo-pumps of the ITER torus, of the neutral beam injectors and of the cryostat together with the corresponding backing pump trains, as well as the provision of a powerful leak localization technique. The TIMO facility operated at the Forschungszentrum Karlsruhe will be employed for full scale cryo-pump tests.

Tritium in first wall materials of fusion machines is a matter of concern not only for inventory purposes but also for safety reasons. In the first wall carbon tiles, tritium is found predominantly in a very thin co-deposited layer and/or implanted in the surface layers. Since the tiles must eventually be disposed off in a repository it is necessary to reduce the total tritium inventory in the tiles down to the corresponding Low Level Waste (LLW) category. In this context, the development of detritiation techniques for graphite and CFC tiles removed from the first wall of fusion machines is of paramount importance. However, the experimental techniques investigated until now are limited by the conditions and techniques permissible inside the torus. Nevertheless, for tiles removed from the reactor more severe treatment techniques are acceptable. Working in that direction, numerous screening tests have been carried out at the Tritium Laboratory in Karlsruhe (TLK).

A new hydrogen isotope separation column has been constructed, manufactured and tested with deuterium and protium. After successful commissioning tests the column was connected to the existing gas chromatographic Isotope Separation System within a glove box. It was demonstrated that equipment which has already been operated with large amounts of tritium can be opened without spreading excessive contamination if proper purging has been carried out before the breach of the primary system. Commissioning with deuterium and small amounts of tritium after the new column was integrated into the existing process circuit in the glove box confirmed the good separation of the new column already demonstrated before.