Volumetric radioassay of lead bricks being considered for unrestricted release

The hardware, software, and a protocol have been developed for the screening of lead bricks for free release or recycle from the Idaho National Engineering Laboratory. The procedure for measuring the background from a sampling of ``clean`` lead bricks and for deducing the decision limits (in pCi/g) have been developed. At the decision limit, a radioactive lead brick would be detected with 95% confidence if it were present. The total and peak efficiencies of a 2.54-cm diameter {times} 2.54-cm high NaI(Tl) scintillation detector for the counting geometry of the present study were measured with a mixed radionuclide standard and calculated with a Monte Carlo program, CYLTRAN. The deduced decision limit for the counting conditions of the present study were 0.0588 pCi/g by analyzing the entire spectrum and 0.256 pCi/g by analyzing the 661-keV peak region for a 900-second count

Converting Simulated Sodium-bearing Waste into a Single Solid Waste Form by Evaporation: Laboratory- and Pilot-Scale Test Results on Recycling Evaporator Overheads

Conversion of Idaho National Engineering and Environmental Laboratory radioactive sodium-bearing waste into a single solid waste form by evaporation was demonstrated in both flask-scale and pilot-scale agitated thin film evaporator tests. A sodium-bearing waste simulant was adjusted to represent an evaporator feed in which the acid from the distillate is concentrated, neutralized, and recycled back through the evaporator. The advantage to this flowsheet is that a single remote-handled transuranic waste form is produced in the evaporator bottoms without the generation of any low-level mixed secondary waste. However, use of a recycle flowsheet in sodium-bearing waste evaporation results in a 50% increase in remote-handled transuranic volume in comparison to a non-recycle flowsheet

INTEC High-Level Waste Studies Universal Solvent Extraction Feasibility Study

This report summarizes a feasibility study that has been conducted on the Universal Solvent Extraction (UNEX) Process for treatment and disposal of 4.3 million liters of INEEL sodium-bearing waste located at the Idaho Nuclear Technology and Engineering Center. This feasibility study covers two scenarios of treatment. The first, the UNEX Process, partitions the Cs/Sr from the SBW and creates remote-handled LLW and contact-handled TRU waste forms. Phase one of this study, covered in the 30% review documents, dealt with defining the processes and defining the major unit operations. The second phase of the project, contained in the 60% review, expanded on the application of the UNEX processes and included facility requirements and definitions. Two facility options were investigated for the UNEX process, resulting in a 2 x 2 matrix of process/facility scenarios as follows: Option A, UNEX at Greenfield Facility, Option B, Modified UNEX at Greenfield Facility, Option C, UNEX at NWCF, th is document, covers life-cycle costs for all options presented along with results and conclusions determined from the study

CsIX/TRU Grout Feasibility Study

A settlement agreement between the Department of Energy (DOE) and the State of Idaho mandates that liquid waste now stored at the Idaho Nuclear Technology Engineering Center (INTEC - formerly the Idaho Chemical Processing Plant, ICPP) will be calcined by the end of year 2012. This study investigates an alternative treatment of the liquid waste that removes undissolved solids (UDS) by filtration and removes cesium by ion exchange followed by cement-based grouting of the remaining liquid into 55-gal drums. Operations are assumed to be from January 2008 through December 2012. The grouted waste will be contact-handled and will be shipped to the Waste Isolation Pilot Plant (WIPP) in New Mexico for disposal. The small volume of secondary wastes such as the filtered solids and cesium sorbent (resin) would remain in storage at the Idaho National Engineering and Environmental Laboratory for treatment and disposal under another project, with an option to dispose of the filtered solids as a r emote-handled waste at WIPP

Research and engineering assessment of biological solubilization of phosphate

This research and engineering assessment examined a microbial phosphate solubilization process as a method of recovering phosphate from phosphorus containing ore compared to the existing wet acid and electric arc methods. A total of 860 microbial isolates, collected from a range of natural environments were tested for their ability to solubilize phosphate from rock phosphate. A bacterium (Pseudomonas cepacia) was selected for extensive characterization and evaluation of the mechanism of phosphate solubilization and of process engineering parameters necessary to recover phosphate from rock phosphate. These studies found that concentration of hydrogen ion and production of organic acids arising from oxidation of the carbon source facilitated microbial solubilization of both pure chemical insoluble phosphate compounds and phosphate rock. Genetic studies found that phosphate solubilization was linked to an enzyme system (glucose dehydrogenase). Process-related studies found that a critical solids density of 1% by weight (ore to liquid) was necessary for optimal solubilization. An engineering analysis evaluated the cost and energy requirements for a 2 million ton per year sized plant, whose size was selected to be comparable to existing wet acid plants

Waste disposal options report. Volume 1

This report summarizes the potential options for the processing and disposal of mixed waste generated by reprocessing spent nuclear fuel at the Idaho Chemical Processing Plant. It compares the proposed waste-immobilization processes, quantifies and characterizes the resulting waste forms, identifies potential disposal sites and their primary acceptance criteria, and addresses disposal issues for hazardous waste