TABLE 5.1 DOE Sodium-bonded Spent Nuclear Fuel Inventory

Fuel Category

Quantitya (MTHM)


Storage Location

EBR-II driver


U metal 10% Zr alloy


EBR-II driver


U metal 5% fission alloyb


EBR-II blanket


DUd metal


Fermi-1 blanket


DUd metal Mo alloy




U metal Mo alloy

Hanford and ANL-W

aPre-demonstration values.

bFission alloy contains Mo, Ru, Rh, Pd, Zr, and Nb.

cIdaho Nuclear Technology and Engineering Center, located at Idaho National Engineering and Environmental Laboratory (INEEL).

dDU = depleted uranium.

eFFTF = Fast Flux Test Facility.

  • Complete the development of the zeolite column to separate plutonium and fission products from the salt; and, finally,

  • Complete the development of the lithium oxide reduction step as a front end-process to treat oxide fuel.

Of these post-demonstration activities, pressureless sintering and HTER and zeolite column development are required only if it could be shown that their implementation would significantly reduce the cost or the time required to treat the remaining sodium-bonded fuel. The remaining activity, development of the lithium oxide reduction step, is required if a decision is made by DOE to treat that fraction of the sodium-bonded EBR-II fuel, referred to as “disrupted” EBR-II fuel, for which the cladding has been breached and deterioration by oxide formation has occurred, or if DOE wishes to develop EMT for processing oxide fuels. Alternatively, the lithium oxide reduction work should continue if DOE wants an electrometallurgical process development that can treat oxide fuels.

In the remainder of this chapter, the committee discusses the post-demonstration activities planned by ANL and offers related recommendations.


The specific data required for waste form qualification are determined by the need to ensure the long-term safety of a deep geologic repository containing such waste forms. DOE-RW is preparing, but has not yet finalized, acceptance criteria for DOE spent nuclear fuel and high level waste.

The technical basis for such acceptance criteria has been addressed in previously published safety assessments for the proposed repository at Yucca Mountain, Nevada. The DOE-RW Yucca Mountain Project has conducted several system studies on repository safety. The total system performance assessment study in 1995 (TSPA-95),4 in particular, reviews the technical basis for data needs with respect to waste-form and repository performance. The TSPA-95 report makes clear that there are several aspects to waste-form performance that assure safe levels for radionuclide releases from a repository.

The first aspect is the waste-matrix dissolution rate, also called the alteration or leach rate, that controls the long-term release of soluble radioactive elements. The second aspect is the solubility-limited concentration for a given radioactive element, imposed either by equilibrium between groundwater and a stable waste form matrix or by equilibrium between groundwater and alteration products that form because they are more stable than the dissolving waste form matrix. These aspects correspond exactly to those identified in a previous NRC study (the


Office of Civilian Radioactive Waste Management, Total System Performance Assessment Viability Assessment, B00000000-01717-4301-00005, U.S. Department of Energy, Washington, D.C., 1995, Chapter 6.

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