Tritium Laboratory Research Articles

The Closed Tritium Cycle of the Tritium Laboratory Karlsruhe

The Tritium Laboratory Karlsruhe (TLK) was commissioned with tritium in 1994 and since then has continuously improved its infrastructure and has expanded its experimental activities. With a fully closed tritium cycle and the license to handle 40 g of tritium TLK is an almost unique experimental facility. More than 10 glove box systems with a total volume of about 125 m3 are operated to house experiments and infrastructure facilities on an area of more than 1000 m2. Today TLK has about 23 g of tritium on site. The paper describes the closed tritium loop of the TLK infrastructure and its links to different experiments. Some experience gained during operation of TLK is also presented.

Decontamination, Dismantling and Refurbishing of the PETRA Glove Box at the Tritium Laboratory, Karlsruhe

The PETRA facility at the Tritium Laboratory Karlsruhe (TLK) has finished its useful life and the glove box and auxiliary systems are being refurbished. During the lifetime of PETRA the glove box became contaminated with a small amount of tritium but the source has not been positively identified. Removing large redundant components would be hazardous as this would require removing the glove box panels and thus exposing the inner surfaces to moist air which would release tritium. Over several months defined amounts of water have been introduced into the glove box daily which has liberated significant quantities of tritium which has subsequently been absorbed by the in-built tritium retention system. This technique has slowly reduced the tritium liberated at each step. The large components, such as a getter bed, catalyst bed and a permeator, have been detritiated as far as possible in-situ in readiness for disposal once it is safe to remove them from the glove box.

Collection and analysis of data related to fusion machines (JET and TLK) operating experience on component failure

This paper presents the summary of the activities related to data collection on the operating experiences gained in the Joint European Torus (JET) for the Vacuum and Active Gas Handling Systems (VS, AGHS) and in the Tritium Laboratory of Karlsruhe (TLK). Failures/malfunctions, including information on failure modes and, where possible, causes and consequences of failures, have been identified, as well as, the whole sets of components, which failures/malfunctions are related to. Reference components installed in the plant/facility have been classified and counted in order to find out their amount, related operating hours and/or related demands to operate. Main reliability parameters associated to the components have been estimated too.

Successful Experimental Verification of the Tokamak Exhaust Processing Concept of ITER with the CAPER Facility

One of the design targets for the Tokamak Exhaust Processing (TEP) system of ITER is not to lose more than 10-5gh-1 into the Normal Vent Detritiation system of the Tritium Plant. The plasma exhaust gas therefore needs to be processed in a way that a tritium removal efficiency of about 108 with respect to the flow rate is achieved. Expressed in terms of tritium concentrations this corresponds to a decontamination from about 130 gm-3 down to about 10-4 gm-3 (about 1 Cim-3 = 3.7*1010 Bqm-3). The three step reference process for the TEP system of ITER is called CAPER and has been developed and realized at the Tritium Laboratory Karlsruhe (TLK). After the successful commissioning of the PERMCAT reactor as the key component of the third step detailed parametric tritium testing of the 3 steps involving the processing of more than 300 g tritium has been carried out and decontamination factors beyond the design requirements have been demonstrated for each process step and for the process as a whole. Not only the decontamination factor of 108 as required by ITER, but also the operational mode of TEP as a waste dump for gases from diverse sources has been experimentally validated with the CAPER facility.

Experimental validation of a method for performance monitoring of the front-end permeators in the TEP system of ITER

The reference process for the Tokamak Exhaust Processing (TEP) system of ITER is called CAPER and comprises three different, consecutive steps to recover hydrogen isotopes at highest purity for direct transfer to the cryogenic isotope separation system (ISS). The CAPER process was developed at the Tritium Laboratory Karlsruhe (TLK) and employs a palladium/silver permeator battery as the 1st step to separate more than 95% of the un-burnt deuterium/tritium fuel from impurities like helium, hydrocarbons and water. These so-called front-end permeators have a capacity of about 80 mol h −1/1 m 2 effective surface area if operated under conditions currently specified for ITER. The front-end permeators of ITER should all the time be operated such that coking of the permeation membranes by hydrocarbon cracking is avoided, since this process lead to a reduction of the effective surface area and therefore to a reduction of the performance of the component. At TLK a method to measure the actual performance of a technical permeator has been developed. This method has been successfully tested with the experimental facility for the demonstration of the CAPER process at TLK and appears feasible for the TEP system of ITER.

Comparison of Tritium Component Failure Rate Data

Published failure rate values from the US Tritium Systems Test Assembly, the Japanese Tritium Process Laboratory, the German Tritium Laboratory Karlsruhe, and the Joint European Torus Active Gas Handling System have been compared. This comparison is on a limited set of components, but there is a good variety of data sets in the comparison. The data compared reasonably well. The most reasonable failure rate values are recommended for use on next generation tritium handling system components, such as those in the tritium plant systems for the International Thermonuclear Experimental Reactor and the tritium fuel systems of inertial fusion facilities, such as the US National Ignition Facility. These data and the comparison results are also shared with the International Energy Agency cooperative task on fusion component failure rate data.

Characterisation of flakes generated in JET after DD and DT plasma operations

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.

Assessment of the heating technique as a possible ex situ detritiation method for carbon wall materials from fusion machines

Detritiation by heat treatment of graphite or carbon fibre composite (CFC) tiles retrieved from the Tokamak Fusion Test Reactor (TFTR) was assessed using the standard and well proved technique of full combustion combined with liquid scintillation analysis. Complete carbon tiles were heated in a oven to 500 °C in air for one hour at Princeton Plasma Physics Laboratory (PPPL) and sent to Tritium Laboratory Karlsruhe (TLK) for surface and depth profile analysis. Coring/full combustion technique performed on the treated and untreated tiles, both graphite and CFC, has shown that only the tritium held close to the surface (few tens μm) is efficiently released by baking in air while the bulk tritium is almost unaffected. Therefore, baking the tile under air, even at 500 °C, does not detritiate the bulk.

Simultaneous tritium and deuterium transfer in a water detritiation CECE facility at TLK

The most suitable process for water detritiation is the combined electrolysis catalytic exchange (CECE) process, a combination of electrolysis and catalytic exchange of hydrogen isotopes in a liquid phase catalytic exchange (LPCE) column. Tritium is recovered in gas phase from tritiated water and then injected in the cryogenic distillation system for further enrichment. At the Tritium Laboratory, Karlsruhe (TLK) a CECE facility is running in order to investigate the simultaneous transfer of tritium and deuterium between gas and water. Experiments are performed to determine the height equivalent of theoretical plate (HETP) and the mass transfer coefficients for deuterium and tritium. The purpose of the experiments is to provide the data required for the design of large water detritiation facility for JET and ITER.

Cold trap and cryogenic molecular sieve adsorber: components for tritium extraction from the purge gas of the HCPB-breeder blanket for ITER

The helium purge gas loop for tritium extraction from the ceramic breeder blanket (Helium Cooled Pebble Bed, HCPB) of ITER (International Thermonuclear Experimental Reactor) includes as main components a liquid-nitrogen cooled cold trap (CT) for the retention of tritiated water produced in the breeder, and a liquid-nitrogen cooled molecular sieve adsorber bed to extract various isotopic species of hydrogen (H 2, T 2, HT) as well as N 2 and O 2. In a preliminary test rig, a CT was operated under once through conditions to obtain data for the final design. The results of humidity measurements confirm the efficiency of the system. To demonstrate hydrogen extraction from a helium gas flow at the operating temperature of 78 K, the test rig will also include a liquid-nitrogen cooled adsorber bed designed and constructed at the Forschungszentrum Karlsruhe. As the next step, a test facility on a technical scale (one-sixth of the flow rate of the ITER-HCPB) will be built in a glovebox in the Tritium Laboratory Karlsruhe (TLK) for tritium experiments with the two main components of the extraction process, i.e. the CT and the molecular sieve bed.

Possible techniques for the detritiation of first wall materials from fusion machines

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 codeposited layer and/or implanted in the surface layers. Since the tiles must eventually be disposed off in a repository it is desirable to reduce the total tritium inventory in the tiles down to the corresponding low level waste 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).

Exchange of tritium implanted into oxide ceramics for protium by exposure to air vapors at room temperature

The exchange of tritium, implanted into an oxide ceramic with a pure tritium ion beam, for protium by exposure to normal air at room temperature has been measured in a collaboration experiment with the Tritium Laboratory group, JAERI. It is shown that the release of THO produced by the exchange H 2O+T→THO+H at the surface of the crystallites decreases monotonically with exposure time, while the release of TH gas increases in the beginning, reaches a maximum and decreases monotonically as the time increases. The latter behavior of TH gas release gives an important evidence for bulk recombination of H uptaken from the surface with T implanted. These results are discussed based on the one-way diffusion model proposed previously for the explanation of the D–H exchange experiment.

Investigation of simultaneous tritium and deuterium transfer in a catalytic isotope exchange column for water detritiation

A catalytic isotope exchange facility for water detritiation is operated at the Tritium Laboratory Karlsruhe (TLK) in order to investigate the simultaneous transfer of tritium and deuterium between various molecular hydrogen isotopes and deuteriated/tritiated water. Four catalyst-packing combinations from different manufacturers (SCK-CEN Belgium, Mendeleev University Russia, Institute of Cryogenics and Isotopic Technologies (ICIT) Romania-TLK, Showa Eng. Japan) are currently being tested. Experiments have been performed to determine the height equivalent of theoretical plate (HETP) and the mass transfer coefficients during deuterium, tritium or simultaneous deuterium and tritium exchange between deuteriated and tritiated water and gaseous hydrogen isotopes. In the simultaneously tritium and deuterium transfer experiments, both the transfer of tritium and deuterium from liquid to gas and the transfer of tritium from liquid to gas and deuterium in the opposite direction from gas to liquid was investigated.

Tritium Processing and Tritium Laboratory in the International Fusion Materials Irradiation Facility (IFMIF)

ABSTRACTThis paper describes design considerations on the tritium processing and the tritium laboratory in the International Fusion Materials Irradiation Facility (IFMIF) that is being developed for neutron irradiation testing of candidate fusion materials. In the IFMIF the tritium is generated by various reactions. However, the irradiated materials will have high activities. The tritium processing method is designed on the basis of the evaluation for tritium source and generation rate. The tritium laboratory for post irradiation examination (PIE) of the activated/tritium contaminated materials is evaluated.

Investigation of Separation Performances of Various Isotope Exchange Catalysts for the Deuterium-Hydrogen System

ABSTRACTAt the Tritium Laboratory of the Forschungszentrum Karlsruhe (TLK) an experimental facility is running which is designed for the intercomparison of several hydrophobic catalysts for use in an liquid phase catalytic exchange (LPCE) column. The catalysts under comparison are from Russia, Belgium and Romania. The intercomparison is being performed by computing the height equivalent of theoretical plates (HETP) and the mass transfer coefficients for HD transfer from gas to water using the measured values of the composition. The range of HD concentration in hydrogen as carrier gas was 1000 ppm up to 2%. The gas and liquid composition at the bottom and at the top of the column and the condensed vapour composition at the top of the column are measured by mass spectrometry and IR spectrometry.

Post Service Examination of Tritium Process Components from the Tritium Laboratory Karlsruhe

ABSTRACTThe Tritium Laboratory Karlsruhe (TLK) is developing processes, devices and measurement techniques for the fuel cycle of fusion reactors. Several tritium processing components from tritium operations of experimental and infrastructure facilities of TLK have reached the end of their useful lives, are no longer useable or are to be reconditioned for future use. These components are in the process of post-service examination to determine the condition of the working materials inside these components and the changes that have resulted from operation with hydrogen isotopes, predominantly tritium. Another important aspect is the preparation of some of these components, particularly metal hydride storage beds, for safe disposal. The results are of importance for the design, operation and particularly decommissioning of components of present and future tritium facilities and processing plants, such as JET and ITER. The activities are ongoing and the results to date are presented together with plans for future work.

Tritium Activities in Germany

ABSTRACTDue to the renunciation of Germany to the possession and fabrication of atomic weapons, no military-related tritium processing or research is undertaken. Tritium work is mainly performed within the framework of the development of nuclear fusion reactors. Focal point of tritium activities in Germany is the Karlsruhe Tritium Laboratory (TLK) at the Forschungszentrum Karlsruhe (FZK), where tritium amounts are processed on a technical level. The TLK is the main European laboratory for nuclear fusion-related work directed to supporting the development of ITER (International Thermonuclear Experimental Reactor) and exploiting JET (Joint European Torus). Moreover at FZK vacuum pumping development and environmental impact of tritium issues are addressed. Apart from the activities at FZK, there are several other sites in Germany, where theoretical and practical work on tritium for special R&D purposes is being performed.

Use of Gas Chromatography in the Tritium Laboratory Karlsruhe

ABSTRACTOne of the objectives of the Tritium Laboratory Karlsruhe (TLK) is to develop new analytical techniques or improve existing ones for the qualitative and quantitative determination of the composition of gas mixtures containing tritium. Among the analytical techniques (mass spectrometry, gas chromatography, laser Raman spectroscopy, calorimetry, ionisation chambers, etc.) available at die TLK, gas chromatography plays a special role because different gas chromatographs are used not only for basic and applied research, but also for infrastructure applications. The TLK gas chromatographs are evaluated on the basis of flow diagrams and typical chromatograms. Further details such as type of columns, detectors and detection limits are also given.

The Behaviour of Zirconium-Cobalt as a Material for Tritium Storage

ABSTRACTThe Tritium Storage facility of the Tritium Laboratory Karlsruhe (TLK), was recently upgraded by four additional uranium beds and one containing ZrCo. Now it is possible to compare under very similar conditions the sorption/desorption properties of uranium and those of ZrCo. To test the adequacy of ZrCo for routine, the getter was folly disproportionated repeatedly using either protium, deuterium or tritium. In all cases was it possible to completely restore the original sorption/desorption properties of the getter. In view of the needs of the Storage and Delivery System (SDS) of the ITER-FEAT Tritium Plant, i.e. delivery of T2 90% - D2 10% and pure D2 to generate mixtures of various compositions, the isotope effects during sorption/desorption cycles of H-D and D-T mixtures in ZrCo granules have been investigated. A transportable storage vessel identical to those containing uranium presently used at TLK for the transport of tritium has been designed, build and filled with ZrCo. These beds fit into the highly sensitive calorimeters available at TLK. With these beds the tritium permanently trapped in U and ZrCo can be compared under similar operation conditions.

Upgrade of the Isotope Separation System at the Tritium Laboratory Karlsruhe

ABSTRACTA 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.