The following HTML text is provided to enhance online
readability. Many aspects of typography translate only awkwardly to HTML.
Please use the page image
as the authoritative form to ensure accuracy.
U.S.-GERMAN COOPERATION IN THE ELIMINATION OF EXCESS WEAPONS PLUTONIUM
C.2 EXPERIENCE WITH VITRIFICATION
In Europe there are two principal vitrification technologies: the French Advanced Vitrification Method (AVM), which has two stages and a production rate of continuously 30–40 liters per hour. The glass is periodically discharged from the melter. The advantages of this method are that it is simple, easy, and cheap; the disadvantages are that it does not have a high capacity and its product quality is low.
The other process is called the liquid-fed ceramic waste glass melter (LFCM). It represents a one-stage technique. The HLLW is directly solidified in a Joule-heated melter without a preceding drying and calcining step. This is one approach to scaling up to high capacities. Several hundred liters per hour seem possible. A “cold cap” acts as barrier against volatiles, and it is discharged periodically. Experience with this process exists in Germany and in Russia.
In the United States, the Savannah River Plant in South Carolina has extensive experience with vitrification. Experimental work has been carried out in which as much as seven percent of plutonium was successfully dissolved in the glass in a bench scale test. The large scale vitrification program has suffered from considerable delays, however, which DOE hopes will be overcome in the near future, and as already mentioned, operation is expected to begin in 1996. A second U.S. plant at Hanford, called the Hanford Waste Vitrification Project, plans to use very large melters that probably would not be suitable for plutonium use.
Since plutonium vitrification on a substantial scale has not been done, more technical uncertainties must be solved than in the case of the MOX option. Several questions must be answered in future engineering work. It is unclear if adding more than 1.2 percent of WPu by weight to existing processes is possible without technical problems. Critically problems could be solved by building smaller melters. The safety of these processes would still have to be certified which would also take some time. A waste form could also be developed specifically for the plutonium disposition mission with a radiation protection barrier from highly radioactive Cesium-137, which is available in the United States. This possibility would still require substantial R&D and costs.
Accumulation of plutonium and release into the environment must be avoided. Licensing of the plutonium-bearing glass as a suitable waste form for emplacement in a geological repository would be the highest hurdle. But assessments show that these obstacles seem to be resolvable, since plutonium leakage is low in comparison to other radioactive isotopes. It is also not fully clear how to assure that the logs remain under-critical over the long term. This problem could probably be solved by adding neutron absorbers with a low diffusion capability to the glass. The effect of this must be demonstrated for licensing. The selective (more rapid) leaching of boron and other absorbers should be addressed.
Russia has experience with vitrification, but largely with phosphate glass rather than borosilicate glass. While there is strong Russian interest in pursuing the use of vitrification further for high-level waste disposal, Russia shows no interest in using vitrification as a means for disposition of WPu. The reason is Russian insistence on putting the energy content of WPu to constructive use for electricity generation. In contrast, there is considerable interest in this method in the United States because this method of disposal would not generate a perceived conflict with the present policy, which does not permit the use of plutonium in the commercial nuclear fuel cycle. Furthermore, it is expected to cost no more than the added cost of WPu disposal via the