First generation fusion machines will burn the hydrogen isotopes deuterium and tritium as fuel. Tritium, which is radioactive, must be handled carefully and any byproducts must be disposed of as radioactive waste. Tritiated waste will fall into one of three categories: gaseous, liquid, or solid. Sources and expected quantities of tritiated waste that are likely to be generated by a fusion machine will be discussed. Strategies, options and processes for dealing with tritium containing wastes will also be presented. Final disposition of each category of tritium contaminated waste is governed by regulations and guidelines. These regulations and guidelines, and how they might apply to a fusion reactor facility, will be outlined.

All fusion reactors that use deuterium-tritide (DT) for fuel will produce tritium-containing water. The quantity and quality (tritium per unit volume) of tritiated water produced depends on several factors. In general, however, the higher the reactor availability the greater the quantity and quality of water produced. This water must be collected and processed to avoid worker exposure and release to the environment. The options for disposition of this water are limited, and in more advanced reactors the tritium contained in water could represent a significant loss to the fuel cycle. The technology currently exists or is being developed to support near term, low availability machines. The technology to support more advanced concepts must be identified and further developed so that it is available when needed.

The purpose of this study was to determine the constraints imposed on the FED/INTOR fusion plants by tritium contamination in the reactor hall. Limitations imposed by gamma radiation levels and the maintenance philosophy adopted were included. The cost (capital and operating) required to maintain 5, 50 and 500 μCi/m was evaluated. Three different maintenance strategies (unsuited worker, suited worker and robotics) were studied to determine if tritium contamination placed a constraint on reactor availability.

The effect of either concrete or stainless steel walls on the cleanup time required to attain tritium (HTO) atmospheric levels ≤ 0.5 mCi/m3 was assessed. Calculations were performed for concrete; experimental results are presented for stainless steel.

The TMIST-2 irradiation experiment was conducted in the Advanced Test Reactor at Idaho National Laboratory to evaluate tritium permeation through Type 316 stainless steel (316 SS). The interior of a 316 SS seamless tube specimen was exposed to a 4He carrier gas mixed with a specified quantity of tritium (T2) to yield partial pressures of 0.1, 5, and 50 Pa at 292 °C and 330 °C. In situ tritium permeation measurements were made by passing a He–Ne sweep gas over the outer surface of the specimen to carry the permeated tritium to a bubbler column for liquid scintillation counting. Results from in situ permeation measurements were compared with predictions based on an ex-reactor permeation correlation in the literature. In situ permeation data were also used to derive an in-reactor permeation correlation as a function of temperature and pressure over the ranges considered in this study. In addition, the triton recoil contribution to tritium permeation, which results from the transmutation of 3He to T, was also evaluated by introducing a 4He carrier gas mixed with 3He at a partial pressure of 1013 Pa at 330 °C. Less than 3% of the tritium resulting from 3He transmutation contributed to tritium permeation.