These and other questions are discussed below under the categories of
zeolite column loading,
glass-bonded zeolite production, and
waste form testing and acceptability.
To date, consideration has been given largely to use of the electrometallurgical technique for disposition of excess weapons plutonium as an add-on to an existing spent nuclear fuel treatment process. For this reason, there is considerable uncertainty about concentrations of fission products, transuranic elements, residual uranium, and added 137Cs in the molten salt feed to the zeolite columns.
Zeolite sorption of molten chloride salts has been under development at ANL for more than 5 years. The chemistry of ion exchange and salt occlusion in the zeolite has been studied in the absence of large quantities of radioactive materials, but the technology has not been demonstrated on a large scale. The current development scale is with batch sizes of about 1 kilogram.
Work has begun on ion exchange column tests using simulated waste. These studies must ultimately provide sufficient engineering data to determine operable ranges of column configuration, ion exchange materials, and ion exchange particle dimensions, as well as effective operating conditions. The tests must also establish the thermal, chemical, mechanical, and radiological stabilities under expected column loading conditions.
Results to date have shown that ions are removed selectively by the zeolite, dependent largely on ionic charge, with higher-valent ions displacing lower-valent ions. Additional important questions remain concerning how plutonium and the other radioelement ions will load spatially on the zeolite columns, the capacity of the columns, and the ability to maintain sufficiently uniform loadings of plutonium and fission products. Information also is needed on the effects of temperature distribution in the zeolite columns on loading selectivity and sorption kinetics of the various radioelements. Finally, some of these effects will introduce complications into criticality control and engineering optimization of zeolite column sizing and performance, as well as the ultimate attainment of “spent fuel equivalency” of the GBZ.
Fabrication of the GBZ waste is being studied at the 10- to 100-gram batch scale. Screening tests are under way to select the best glass frit to produce an immobilization form having the necessary leach resistance and mechanical properties. Much of the development equipment became available only in late 1995. The committee is concerned with the potentially long development period that may be necessary for the iterative testing and optimization of the GBZ fabrication process.
The GBZ characteristics will determine the actual mechanism (e.g., ion exchange, dissolution) for release of radioelements from the zeolite waste form over the long time periods relevant to geologic disposal. The committee has several areas of concern regarding the progress and adequacy of the characterization and performance testing of the GBZ waste form. In addition, changes in the zeolite structure could alter the suitability of GBZ as a long-term waste form.