An Evaluation of the Electrometallurgical Approach for Treatment of Excess Weapons Plutonium (1996) / Chapter Skim
Currently Skimming:
B Argonne National Laboratory Responses to Questions Submitted on November 3, 1995, by the Committee
B Argonne National Laboratory Responses to Questions Submitted on November 3, 1995, by the Committee
Pages 39-53
The Chapter Skim interface presents what we've algorithmically identified as the most significant single chunk of text within every page in the chapter.
Select key terms on the right to highlight them within pages of the chapter.
From page 39... ...
The waste salt input shown in the flow sheet is actually within the same electroref~ner vessel. The important features of this waste salt are the fission products from the spent fuel that contribute to the radiation source for meeting the "spent fuel standard", and the UCI3 that helps convert metallic plutonium to PuCI3 for ion exchange with the zeolite.
|
From page 40... ...
Approximately 1000 such canisters will be needed to dispose of the 50 MT weapons-grade plutonium, so a total 137CS activity of 20 MCi will be needed. The gamma radiation from the spent fuel fission products could be used to reduce the number of cesium capsules needed, to reduce the amount of plutonium in each canister (by increasing the number of canisters)
|
From page 41... ...
We believe that the metal and mineral waste forms must be fully qualified for repository disposal, just as the spent fuel and glass waste forms must be. The Waste Acceptance Systems Requirements Document identifies glass and spent fuel as "standard" waste forms.
|
From page 42... ...
There are differences in fission product distribution between fast and thermal reactor spent fuels, but we have not seen any differences that would affect the development program. We have therefore spent our limited resources only on a detailed breakdown for the EBR-II fuel.
|
From page 43... ...
of major and minor components in zeolite-A before and after contact with molten salt? The following table provides the compositions of a typical salt-occlu(led zeolite A and "anhydrous" zeolite A
|
From page 44... ...
We have an isothermal equilibrium model that indicates that the distribution evolves as one would expect as more and more salt is passed through the column-"waves" of ions pass down the column as each ion is displaced by more strongly-retained ions until eventually the column would become entirely uniform and at equilibrium with the inlet salt composition. Unfortunately, some of the kinetics are slow and, of course, the fission products generate heat in the column.
|
From page 45... ...
Calculations of heat generation have not been done specifically for the Pu disposition forms; however, calculations were done for heat generation by fission products and transuranic elements in the glass-bonded zeolite waste form for disposal of spent fuel wastes. The results of these calculations were published in the ANE Report, ANL-IFR-165, June 1992, for the calcium reduction system, and the calculations were updated in ANL-IFR-253, March 1995, for the lithium reduction system.
|
From page 46... ...
Excessive volumetric heat generation is not a problem with the EBR-II fuel because of its age. This applies as well to all of the spent fuel in the DOE inventory.
|
From page 47... ...
The as-synthesized powcler particles are typically in the 1-10 micron diameter size range and each powcler particle consists of an intergrown mass of several crystals. (cJ For co;lumn ion exchange or adsorption applications these powder particles are bonded, often with clay binders (typically with about 80 wt zeolite and 20 wt binder, on an anhydrous weight basisJ to form larger particles, e.g., 20 x 50 mesh particles for ion exchange, or ~ x 12 mesh or 1/16" pellets for gas phase adsorption applications.
|
From page 48... ...
That work has been de-emphasized due to funding constraints. As far as using sodalite as an ion exchanger, we did one or two experiments exposing large natural sodalite specimens to molten salt.
|
From page 49... ...
For the electrometallurgical treatment of DOE spent nuclear fuel, we need to have (preferably one) material to retain all the fission products, not just cesium, and we strongly prefer a material which will remove all of these fission products from the electrolyte salt.
|
From page 50... ...
What literature searches, reviews, or reports are available on molten salt ion exchange of zeolites? (The committee ?
|
From page 51... ...
1973. Ion exchange in molten salts-VI: The occluded Sodium Nitrate in Zeolite A as an Anion exchanger.
|
From page 52... ...
1974. Surface phenomena on Zeolite A in molten salt media-I: The mechanism of bivalent cation exchange through metal-nitrate complex formation in the zeolite cages.
|
From page 53... ...
Appendix B.: ANT Responses to Questions (November 3, 1995)
|
Key Terms
This material may be derived from roughly machine-read images, and so is provided only to facilitate research.
More
information on Chapter Skim is available.