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Treatment of sodium-bearing waste (SBW) at the Idaho Nuclear Technology and Engineering Center (INTEC) within the Idaho National Engineering and Environmental Laboratory is mandated by a Settlement Agreement between the Department of Energy and the State of Idaho. One of the requirements of the Settlement Agreement is to complete treatment of SBW by December 31, 2012. To support both design and development studies for the SBW treatment process, detailed feed compositions are needed. This report contains the expected compositions of these feed streams and the sources and methods used in obtaining these compositions. iv SUMMARYA sodium-bearing waste (SBW) treatment facility will treat liquid wastes contained in existing and new tanks at the Idaho Nuclear Technology and Engineering Center (INTEC). Unless removed before treatment, a small amount of solids will be entrained in these liquid feed streams. The treatment facility may also treat tank heel sludges that remain in the tanks after the liquids are withdrawn.This document provides the most recent compilation of the volumes and compositions of these feed streams. As new characterization data are received and as changes are made in the INTEC Tank Farm management plans, this document will be updated. The assumptions and source documents used in calculating the treatment process feed compositions are identified in this report.Two treatment processes are being considered for treatment of SBW. One process, referred to as the "CsIX process," removes cesium from the liquid waste and grouts the cesium-free liquid. For the CsIX process, suspended and heel solids would likely be separated from the feed streams and treated in a separate process. The second process, referred to as "direct vitrification," treats both liquids and solids from the INTEC tanks.Current Tank Farm management plans show that either facility would be required to treat six separate feed streams. Three of these feed streams are "SBW" -acidic, radioactive, and hazardous liquid waste containing small amounts of undissolved solids. SBW has been generated mostly from past decontamination activities at the INEEL. Another feed is a high-solids sludge from heels in existing tanks. The final two feeds are mostly liquid wastes from future operations at the INEEL, often referred to as "newly generated liquid waste" (NGLW). These NGLW streams may be similar in composition to SBW, but insufficient information is available from which to project NGLW compositions. Thus, this report contains composition data for the three SBW feeds and the heel solids but not for the future NGLW feeds.Less data, and hence more uncertainty, is present in estimates of solid compositions and quantities than liquid compositions and quantities.

A major element of the Next Generation Nuclear Plant (NGNP)/Advanced Gas Reactor (AGR) Fuel Development and Qualification Program is developing fuel fabrication processes to produce high quality uranium-containing fuel kernels, TRISO-coated particles and fuel compacts needed for planned irradiation tests. The goals of the program also include developing the fabrication technology to mass produce this fuel at low cost. Kernels for the first AGR test, AGR-1, consisted of uranium oxycarbide (UCO) microspheres that were produced by an internal gelation process followed by high temperature steps to convert the UO 3 + C "green" microspheres to UO 2 + UC x . The high temperature steps also densified the kernels.Babcock and Wilcox (B&W) fabricated UCO kernels in their Lynchburg facility for the AGR-1 irradiation experiment, which went into the Advanced Test Reactor (ATR) at Idaho National Laboratory in December 2006. An evaluation of the kernel process prior and after these kernels were produced led to several recommendations to improve the fabrication process. These recommendations included testing alternative methods of dispersing carbon during broth preparation, evaluating the method of broth mixing, optimizing the broth chemistry, optimizing sintering conditions, and demonstrating fabrication of larger diameter UCO kernels needed for the second AGR irradiation test, AGR-2.Based on these recommendations and requirements, a test program was defined and performed. Certain portions of the test program were performed by Oak Ridge National Laboratory (ORNL), while tests at larger scale were performed by B&W. The tests at B&W have demonstrated improvements in both kernel properties and process operation. Changes in the form of carbon black used and the method of mixing the carbon prior to forming kernels led to improvements in the phase distribution in the sintered kernels, greater consistency in kernel properties, a reduction in forming run time, and simplifications to the forming process. Process parameter variation tests in both forming and sintering steps led to an increased understanding of the acceptable ranges for process parameters and additional reduction in required operating times. Another result of this test program was to double the kernel production rate. Following the development tests, approximately 40 kg of natural uranium UCO kernels have been produced for use in coater scale up tests, and approximately 10 kg of low enriched uranium UCO kernels for use in the AGR-2 experiment.

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