Spent fuel elements contain appreciable amounts of tritium, partly produced by fission, partly by other nuclear reactions. About half of the tritium is released from the fuel upon dissolution. The rest is bound to the zircaloy of the hulls and is disposed of with them. The fraction of tritium that is released exchanges with water, forming НТО. The total annual input of tritium in a 1400 MT/year reprocessing plant is about 106 Ci. In West Germany a reprocessing plant of this size is supposed to retain 75 to 80 percent.

The fundamental problem of tritium waste management is that there is no simple way to reduce the volume of tritiated water. There are techniques available to minimize the volume generated in reprocessing, e. g., reuse of tritiated water to feed steam jets, and strict confinement of tritium in the first extraction cycle. These techniques, however, add complica­tions to the process. If, therefore, an inexpensive way were available to dispose of untreated tritiated water, severe generation restrictions would not be appropriate. If, however, expensive methods were to be applied, such as solidification or even concentration by isotopic enrichment, the volume generated has to be limited as much as possible.

Another approach is a suitable head-end process in the reprocessing plant, such as voloxidation (Chap. 10, Sec. 4.3). However, such a head-end process is not yet available technology but requires several more years of development.

There are minor quantities of tritium smeared out over the whole reprocessing flow scheme that will ultimately arise as low-activity condensate with tritium concentrations of the order of 10"4 Сі/liter and 1СГ8 Ci./liter of other radionuclides. It is very likely that this can be released to surface waters.

Basically three options are considered to dispose of tritiated water that is stored in tanks and cannot be released.

Deep well disposal. Injection of tritium-containing liquid into isolated aquifers or depleted oil horizons is the most interesting option. This technique has been used increasingly for almost 20 years to dispose of industrial wastes. In the United States, for instance, some hundred injection wells have been drilled and are actually in operation at depths between 60 m and 3600 m. Although there are still licensing problems, this is a safe and economic way to dispose of tritiated water.

This technique will be tested for tritiated water in the neighborhood of the Karlsruhe Nuclear Research Center in West Germany. An isolated oil lens that is exhausted but located in an oil field still being exploited will be used. Thereby any migrations occurring deep underground will be detected.

Solidification. In principle, any solid that contains firmly bound water may be suitable as a solidification form for НТО. This includes drying agents, such as silica gel, molecular sieves, and calcium sulfate, as well as hydraulic cement and organic polymers. Most experience is available with cement, which has been used to solidify non-high-level waste for quite a while.

Although concrete is a monolithic solid, it is quite porous. In contact with water, about a third of the tritium will be released, mainly by isotopic exchange, in the first month. The release may be retarded by coating the cement. Because of the relatively high leachability, cemented НТО would have to be stored in gastight steel cylinders, probably in a nonaccessible geologic repository.

If it turns out that a more leach-resistant and probably more expensive solidification product has to be developed, it may well become beneficial not only to restrict the volume arising from reprocessing but also to further reduce it by isotopic enrichment prior to solidification. An enrichment process suitable for this purpose must provide a very effectively depleted waste stream.

Ocean disposal. In view of the relatively short half-life of tritium and of the enormous isotopic dilution, sea disposal is another alternative for dealing with tritium waste. Transport will be an economic drawback of this alternative, and political and administrative problems will have to be solved.

3.1 1291

All iodine isotopes except 1291 will have decayed prior to reprocessing as long as a large backlog of unreprocessed spent fuel exists. The 1291 activity per metric ton of heavy metal (30,000 MWd/MT) is only 34 mCi. However, its extremely long half-life of 17 million years makes 1291 a permanent contaminant if released to the atmosphere. In shorter-cooled fuel elements radioactive 1311 will also be present and must be recovered.

Practically all iodine from spent fuel will be released upon dissolution with the dissolver off-gas. There are several scrubbing techniques that remove iodine effectively from the off-gas but do not yield a stable product for long-term disposal.

For permanent fixation of 1291 adsorption on silver-loaded adsorbents, such as zeolites, silica, or alumina, will be the choice [PI, W2]. The process is simple, the bed temperature may be relatively high, the product is a dry solid, the chemisorbed iodine is highly insoluble, and the adsorbent is very efficient in removing both organic and inorganic iodine from gas streams.

The 1291 content of spent fuel with an average bumup of 30,000 MWd/MT heavy metal is 211 g/MT corresponding to 34 mCi. This corresponds to an annual production from a 1400 MT/year reprocessing plant of 300 kg 129I. As there will be some isotopic dilution, an iodine-recovery system could conceivably be required to remove 600 kg of iodine annually. If iodine will be adsorbed on silver zeolite beds ready for final disposal, the total amount of iodine waste is then estimated to be about 5 m3/year with a total activity of 50 Ci. The amount of silver corresponding to 600 kg iodine is about 500 kg. Even though there will be excess silver required, this does not seem an unreasonable silv r consumption in view of the overall reprocessing costs. The world’s silver production was almost 104 tons in 1976. There is, however, some research in progress on regeneration of iodine-loaded silver zeolite and reloading the iodine on a lead zeolite.