Appendix C:

WPu Disposition Through Vitrification with HLW

C.1 TECHNOLOGY

The spent fuel standard for the disposal of WPu can also be met when the plutonium is mixed with high-level nuclear waste (HLW) that is already being prepared for disposal and then melted into glass logs. Today that vitrification process is well advanced and is considered to be suitable to convert high-level waste, and in particular high-level liquid waste (HLLW) into a stabilized waste form. The technology has been developed and practiced for over 20 years. There are plants in operation worldwide, including those in Sellafield, The Hague, Mol, Marcoule, Chelyabinsk, and Tokai-Mura, and the U.S. facility at Savannah River is expected to begin operation in 1996.

Various forms of glass can be used for vitrification of HLW with or without the addition of WPu. In the United States borosilicate glass has been the primary choice; the boron in the glass is a useful component to lower criticality risks, and the glass is heat resistant.

For the purpose at hand, the final product of the vitrification process would be a glass log in which the plutonium and the other radioactive wastes are dissolved or suspended, cast into a cylindrical stainless steel shell. The final glass log can be made heavier than one ton, although smaller pieces can of course be fabricated. In principle, plutonium oxide could be dissolved in glass without the addition of HLW and thus be immobilized, but this product would not meet the spent fuel standard and so will not be considered here.

The basic device for vitrifying HLW is a melter into which a glass powder, known as glass frit, is continuously fed along with whatever other material is to be dissolved in the glass. The glass is heated and kept in a molten state, either by joule heating1in the ceramic melter (or in a cooled metal melter) or alternatively, if a metal melter is used, by inductive heating of the melter shell. The molten material can be extracted continuously and cast into suitable molds. Currently the melters that are used for vitrification are quite large and this in turn may pose criticality hazards if too much plutonium is added. Generally the material in the melter is not stirred so that selective concentration of the additives during the melting process cannot be excluded.

At this time the maximum amount of plutonium that can be added to the vitrified glass logs has not been well established. The limit is presumably set both by criticality concerns as well as by the basic solubility of the plutonium in the glass. However, any amount above one percent could dispose of fifty tons of WPu in the United States without increasing the total volume of glass logs which would have to be produced at any rate for disposition of HLW.

1  

Joule heating means passing a strong electric current directly through the molten glass.



The National Academies | 500 Fifth St. N.W. | Washington, D.C. 20001
Copyright © National Academy of Sciences. All rights reserved.
Terms of Use and Privacy Statement



Below are the first 10 and last 10 pages of uncorrected machine-read text (when available) of this chapter, followed by the top 30 algorithmically extracted key phrases from the chapter as a whole.
Intended to provide our own search engines and external engines with highly rich, chapter-representative searchable text on the opening pages of each chapter. Because it is UNCORRECTED material, please consider the following text as a useful but insufficient proxy for the authoritative book pages.

Do not use for reproduction, copying, pasting, or reading; exclusively for search engines.

OCR for page 70
U.S.-GERMAN COOPERATION IN THE ELIMINATION OF EXCESS WEAPONS PLUTONIUM Appendix C: WPu Disposition Through Vitrification with HLW C.1 TECHNOLOGY The spent fuel standard for the disposal of WPu can also be met when the plutonium is mixed with high-level nuclear waste (HLW) that is already being prepared for disposal and then melted into glass logs. Today that vitrification process is well advanced and is considered to be suitable to convert high-level waste, and in particular high-level liquid waste (HLLW) into a stabilized waste form. The technology has been developed and practiced for over 20 years. There are plants in operation worldwide, including those in Sellafield, The Hague, Mol, Marcoule, Chelyabinsk, and Tokai-Mura, and the U.S. facility at Savannah River is expected to begin operation in 1996. Various forms of glass can be used for vitrification of HLW with or without the addition of WPu. In the United States borosilicate glass has been the primary choice; the boron in the glass is a useful component to lower criticality risks, and the glass is heat resistant. For the purpose at hand, the final product of the vitrification process would be a glass log in which the plutonium and the other radioactive wastes are dissolved or suspended, cast into a cylindrical stainless steel shell. The final glass log can be made heavier than one ton, although smaller pieces can of course be fabricated. In principle, plutonium oxide could be dissolved in glass without the addition of HLW and thus be immobilized, but this product would not meet the spent fuel standard and so will not be considered here. The basic device for vitrifying HLW is a melter into which a glass powder, known as glass frit, is continuously fed along with whatever other material is to be dissolved in the glass. The glass is heated and kept in a molten state, either by joule heating1in the ceramic melter (or in a cooled metal melter) or alternatively, if a metal melter is used, by inductive heating of the melter shell. The molten material can be extracted continuously and cast into suitable molds. Currently the melters that are used for vitrification are quite large and this in turn may pose criticality hazards if too much plutonium is added. Generally the material in the melter is not stirred so that selective concentration of the additives during the melting process cannot be excluded. At this time the maximum amount of plutonium that can be added to the vitrified glass logs has not been well established. The limit is presumably set both by criticality concerns as well as by the basic solubility of the plutonium in the glass. However, any amount above one percent could dispose of fifty tons of WPu in the United States without increasing the total volume of glass logs which would have to be produced at any rate for disposition of HLW. 1   Joule heating means passing a strong electric current directly through the molten glass.

OCR for page 70
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

OCR for page 70
U.S.-GERMAN COOPERATION IN THE ELIMINATION OF EXCESS WEAPONS PLUTONIUM MOX route. It is therefore believed that licensing problems in the United States for the glass vitrification disposal option would be less than those faced if MOX fuel is to be burned in commercial reactors.