The aqueous raffinate from the first extraction cycle of light-water reactor (LWR) fuel reprocessing has an original volume of up to 5 m3/MT^ of heavy metal. It is concentrated by evaporation, and the residues of the evaporated raffinates from further extraction cycles may be combined with the concentrate. The result of these operations is the HLW concentrate that will be transferred to the waste management section of the reprocessing plant.

Volume. The volume reduction factor that can be achieved depends strongly on the bumup and on the cooling time the fuel has experienced. According to the length of the cooling period, either the heat generation of the resulting HLW or its content of dissolved solids including fission products and process chemicals are the factors limiting the degree of concentration. Most of the volume reduction is achieved in the HLW evaporator. Some further

11 MT = 1 metric ton = 1 megagram (1 Mg).

evaporation usually takes place in the storage tank, where the temperature of the solution is kept below 60° C. According to the wide range of parameters the specific volumes of the HLW concentrate may range from 0.4 to more than 1 m3 /МТ of heavy metal reprocessed. A value of

0. 6 m3/MT may be used in the design of reprocessing plants to calculate tank volume requirements.

Chemical composition. In a well-designed reprocessing scheme the amount of process chemicals that would appear in wastes will be kept as small as possible. Then the bulk of dissolved solids includes fission products, uranium and plutonium losses, neptunium, and the transplutonium elements. Table 11.2 shows the amounts of fission-product elements (more than 0.1 percent contribution) present in the waste from reprocessing 1 MT of heavy metal from spent LWR fuel with 30,000 MWd/MT bumup, at the time of and 6 years after discharge from reprocessing. According to widely used design parameters, reprocessing is assumed to take place 150 days after reactor discharge. An aging time of 6 years can presently be envisaged prior to solidification of the liquid HLW. In practice, spent fuel will be aged much longer, and storage of liquid HLW may be shorter.

With a specific HLW concentrate volume of 600 liters/MT of heavy metal, the total fission product concentration will be on the order of 50 g/liter and the actinide concentration on the order of 10 g/liter.

Table 11.2 Amounts of fission-product elements and actinide elements in the waste from 1 MT LWR uranium fuel (30,000 MWd/MT bumup) at discharge from reprocessing (150 days cooled fuel elements) and 6 years after discharge (contributions of more than 0.1 percent) according to ORIGEN

Corrosion products play a minor role and make up to about 1 percent of the total solid content of the HLW solution. If gadolinium is used as homogeneous poison for criticality control, it will significantly increase the total solid content of the waste.

The optimum HN03 concentration for tank storage is in the range of 2 to 4 M, determined by the corrosion behavior of stainless steel. The aqueous raffinate stream (HAW) leaves the extraction process with 1 to З M HN03. A good deal of the nitric acid is stripped in the HAW evaporator or destroyed by a reductant (Chap. 10), and more by subsequent water vapor distillation. The final adjustment may be performed in the storage tank, taking advantage of radiolysis, which removes HN03 effectively when the radiolytic products are swept by air.

Table 11.3 shows typical HAW concentrate data as they have been observed in pilot-plant operations and as they are envisaged for commercial operation.

Radioactivity and heat generation. The total radioactivity and the heat generation of the waste solution up to 100 years is essentially a function of the fission-product concentration and of their age. Then the actinides start to contribute significantly. The computer program ORIGEN [B2] permits the calculation of all relevant data such as radionuclide activities, ingestion hazards, element concentrations, heat generation, and neutron generation for fission products and actinides as a function of age. Figure 11.1 shows radioactivity concentrations and heat generation of the waste from 1 MT of heavy metal with 30,000 MWd/MT burnup up to an age of 106 years. Typical maximum radioactivity concentrations and specific heat generations for a freshly filled tank and reprocessing of 150-day-old fuel are of the order of 103 Ci and 10 W/liter, respectively. After about 500 years radioactivity and heat generation have decayed by a factor of more than 1000. The heat generation will then be insignificant.

Table 11.4 shows the contributions (more than 0.1 percent) of individual fission-product nuclides to the total fission-product activity and of individual actinide nuclides to the total actinide activity. After 6 years only nine fission products or fission-product mother/daughter pairs contribute significantly. After 100 years, 90 Sr/90 Y and 137Cs/l37mBa make up to 98 percent of the fission-product activity, and among the actinides 238 Pu, 241 Am, 243Am/239Np, and 244 Cm are responsible for 90 percent of the total alpha radioactivity.

Hazard indices. The radioactivity of the waste is no direct measure of its radiotoxicity or its hazard. When we assume ingestion as the most likely path of incorporating radioactivity from

waste, we may characterize the hazard by the ingestion hazard index. This is defined for an individual radionuclide as radioactivity divided by the radioactivity concentration limit for drinking water (general public) and has the dimension of a volume. It may be understood as the volume of water required to dilute a given quantity of radioactive material so that drinkable water will be obtained. For a mixture of radionuclides, such as radioactive waste, the ingestion hazard index is the sum of the ingestion hazard indices of all radionuclides present. The hazard index characterizes the potential hazard rather than the actual risk associated with the waste. It does not give credit for the various barriers between waste and humans.

Figure 11.2 shows the ingestion hazard index of HLW as a function of time up to 10® years. After about 500 years the actinide radiotoxicity clearly dominates.

HLW from advanced fuel cycles. Advanced fuel cycles that will be considered are plutonium recycling and the LMFBR fuel cycle. There is little difference from LWR waste as far as fission products are concerned. Actinide concentrations, however, are considerably different, as shown in Chap. 8 for spent fuel.

To discuss other properties of advanced fuel-cycle waste, such as chemical composition, is of little use. As yet, the reprocessing technology for advanced fuel cycles is in a very premature state of development.

□adding hulk. Cladding hulls as collected from the chop-leach head end are radioactive due to activation products in the zircaloy and to fission products and actinides from (U, Pu)02 adsorbed at the inside of the hulls. The principal activation products are 60 Co, 125Sb/12SmTe, and 63Ni. Their total activity is on the order of 103 AiCi/g zircaloy after 6 years and about 100 times lower after 100 years. The fission-product activity after 6 years is of the same order of magnitude and is dominated by 137Cs, 90Sr, and tritium. The residual (U, Pu)02 after leaching is estimated to be of the order of 0.1 percent of the charge.

A number of processes are under development for consolidation and volume reduction of the hulls. Melt densification has received much effort, resulting in volume reductions by a factor of 6. Other processes seriously considered are mechanical compaction and consolidation in concrete. In any case, final disposal of consolidated hulls will be similar to that of other HLW. Thus, hulls (and dissolver sludge) will add considerably to the volume of waste eventually to be disposed of with essentially the HLW technology.

2.1 Projections of HLW Generation

Projections of future waste generation depend strongly on projections of nuclear power generation and of reprocessing capacity. Therefore they have a high degree of uncertainty.

Nudear capacity forecasts have considerably decreased over the last years, and it is presently an open question when reprocessing plants will go on stream. Table 11.5 shows the most recent estimate of the amount of solidified HLW to be accumulated at a federal repository in the United States [B4]. There may be a shift of time scale, but this will not greatly affect the general conclusion from this table: Early in the next century on the order of 104 m3 of solidified waste with on the order of 103 MT of actinides and on the order of 1010 Ci of total radioactivity will be collected in the repository and probably be buried underneath the United States.

To put these numbers into perspective, they should be compared with natural radioactivity already contained in the crust of the earth. We consider a layer of soil all over the United States that is 1 m thick. Its volume is of the order of 1013 m3. With an average uranium concentration of 3 ppm, its uranium content is very roughly З X 107 MT, corresponding to a total radioactivity of 1010 Ci. This corresponds to about 1500 MT of J34U in the 1-m layer, a nuclide whose relative ingestion hazard resembles that of 239 Pu within one order of magnitude.

This simple calculation shows that a 1-m layer of the United States contains about as much long-lived radioactivity and actinides as nuclear industry will put beneath several hundred of those layers within the next 30 years. This illustration does not mean that HLW represents no serious and long-lasting hazard potential, but it emphasizes that the amounts of radioactive material dealt with in waste management are not at all alien to nature.