[en] In December 1993 the high power D-T experimental program on the Tokamak Fusion Test Reactor (TFTR) began. The transit the TFTR from a DOE general use facility to a low hazard category III nuclear facility has been completed successfully. The low hazard nuclear facility designation that the allowable on-site tritium inventory not exceed 50,000 Curies (1 Ci = 37 GBq). This is a TFTR Technical Safety Requirement. Tritium sealed in approved shipping containers does riot count against this inventory limit A second Technical Safety Requirement at TFTR is to have no more than 25,000 Ci at risk in a single location. From December, 1993 through mid-August, 1994 about 20 grams of tritium have been used in two gas injector assemblies and twelve neutral beam tritium injectors. The gas injected into TFTR vacuum is pumped by helium cryo-panels in the four neutral beam boxes. During non-operating periods the cryo-panels are warmed and the hydrogen am released and pumped into gas holding tanks in the tritium area. Gas in the holding tanks is oxidized in the Torus Cleanup System (TCS) and the hydrogen isotopes are collected, as water, on disposable molecular sieve beds (DMSB). These beds are then removed from the system and shipped off-site for tritium recovery or for long-term storage. Several problems in the tritium cleanup systems have occurred following a leak of sulfur hexafluoride (SF₆) from a neutral hewn high voltage enclosure ion source and subsequent pumping to the gas holding tanks. These problems included failure of several-moisture sensors, false readings on tritium monitors and, partial loss of catalytic activity in the TCS recombiner. Procedures for dealing with and removing this contaminant gas had to be developed and implemented. The results from this occurrence provide valuable guidance for future tritium burning fusion machines

No abstract available

[en] A small fusion fuel cleanup system has been designed and experimentally tested at CRNL's Tritium Laboratory. The background requirements of fusion fuel systems, in general, are reviewed and the system built at CRNL described. In addition, the contributions of tritium technology developed as part of the CANDU system to the tritium processing requirements of fusion reactors is briefly mentioned

[en] We provide a high-level description of the integrated solution offered by Kurion to extract and safely isolate tritium from any inventory of tritiated water. The Kurion modular detritiation system (MDS) is based on the Combined Electrolysis and Catalytic Exchange (CECE) technology. The MDS can recover more than 99% of the tritium in the form of a metal hydride. The number of MDS can be adjusted to process the entire tritiated water inventory in a preset amount of time

[en] One of the main objectives of the International Thermonuclear Experimental Reactor (ITER) is the testing of tritium breeding blankets for a subsequent DEMO power plant. This paper reviews the main features of the ITER device which are relevant for this mission. Space and remote handling provisions are addressed, and the main test objectives are described. The operation conditions of ITER are outlined, and commented in view of their relevance for blanket testing. While the pulse characteristics are fully appropriate, the wall loading, and in particular the neutron fluence provided by ITER, still require substantial extrapolation to DEMO conditions. These restrictions, together with the envisaged operation schedule, makes careful planning necessary to satisfy the testing intentions of the four ITER Home Teams. (orig.)

[en] The main objective of the described research project is to reduce quantities of tritiated waste and tritium discharges by treatments that allow the trapping of all tritium as tritiated water. This tritiated water could be conditioned or its tritium content could be recovered through recycling. The experimental method that is investigated is a complete two-stage combustion with thermal and catalytic oxidation of the organic liquid into tritiated water for further treatment and tritium free off gases for discharge. For this purpose, an installation capable of accepting liquid organic flows of different origins and compositions (with activities up to 0.2 TBq/l) has been built

[en] Results of an experimental study on a flowing bed process for continuous hydrogen isotope separation are presented. Separation performance was low with a 25% by weight palladium on alumina adsorbent, resulting in a high tritium inventory. In addition, significant breakdown of the solid adsorbent occurred as it recirculated through the process equipment and the product streams were contaminated by the adsorbent carrier gas. Due to these problems, this flowing bed process is predicted to be uneconomic for a full scale plant

[en] The present analyses on the possibilities of extracting tritium from the liquid and solid fusion reactor blankets show up many problems. A consistent ensemble of materials and devices for extracting the heat and the tritium has not yet been integrated in a fusion reactor blanket project. The dimensioning of the many pipes required for shifting the tritium can only be done very approximately and the volume taken up by the blanket is difficult to evaluate, etc. The utilization of present data leads to over-dimensioning the installations by prudence and perhaps rejecting the best solutions. In order to measure the parameters of the most promising materials, work must be carried out on well defined samples and not only determine the base physical-chemical coefficients, such as thermal conductivity, scattering coefficients, Sievert parameters, but also the kinetic parameters conventional in chemical engineering, such as the hourly space rates of degassing. It is also necessary to perform long duration experiments under radiation and at operating temperatures, or above, in order to study the ageing of the bodies employed

[en] Tritium release from the candidate ceramic materials, Li₂O, LiA10₂, Li₂SiO₃, Li₄SiO₄ and Li₂ZrO₃, is being investigated in many blanket programs. Factors that affect tritium release from the ceramic into the helium sweep gas stream include operating temperature, ceramic microstructure, tritium transport and solubility in the solid. A review is presented of the material properties studied and of the irradiation programs and the results are summarized. The ceramic breeder blanket concept is briefly reviewed

[en] This paper surveys the theoretical and experimental results obtained in the field of recovery and enrichment of tritium in Los Alamos National Laboratory, USA, and in Japan Atomic Energy Research Institute, Tokai-mura, Japan.(author)