Rutledge, V. J. | Engineering Papers

DOE OSTI2015

Microfluidic-Based sample chips for radioactive solutions

Historical nuclear fuel cycle process sampling techniques required sample volumes ranging in the tens of milliliters. The radiation levels experienced by analytical personnel and equipment, in addition to the waste volumes generated from analysis of these samples, have been significant. These sample volumes also impacted accountability inventories of required analytes during process operations. To mitigate radiation dose and other issues associated with the historically larger sample volumes, a microcapillary sample chip was chosen for further investigation. The ability to obtain microliter volume samples coupled with a remote automated means of sample loading, tracking, and transporting to the analytical instrument would greatly improve analytical efficiency while reducing both personnel exposure and radioactive waste volumes. Sample chip testing was completed to determine the accuracy, repeatability, and issues associated with the use of microfluidic sample chips used to supply µL sample volumes of lanthanide analytes dissolved in nitric acid for introduction to an analytical instrument for elemental analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2014

Microfluidic-Based Sample Chips for Radioactive Solutions

Historical nuclear fuel cycle process sampling techniques required sample volumes ranging in the tens of milliliters. The radiation levels experienced by analytical personnel and equipment, in addition to the waste volumes generated from analysis of these samples, have been significant. These sample volumes also impacted accountability inventories of required analytes during process operations. To mitigate radiation dose and other issues associated with the historically larger sample volumes, a microcapillary sample chip was chosen for further investigation. The ability to obtain microliter volume samples coupled with a remote automated means of sample loading, tracking, and transporting to the analytical instrument would greatly improve analytical efficiency while reducing both personnel exposure and radioactive waste volumes. Sample chip testing was completed to determine the accuracy, repeatability, and issues associated with the use of microfluidic sample chips used to supply µL sample volumes of lanthanide analytes dissolved in nitric acid for introduction to an analytical instrument for elemental analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2013

The production of advanced glass ceramic HLW forms using cold crucible induction melter

Cold Crucible Induction Melters (CCIM) will favorably change how High-Level radioactive Waste (from nuclear fuel recovery) is treated in a near future. Unlike the existing Joule-Heated Melters (JHM) currently in operation for the glass-based immobilization of High-Level Waste (HLW), CCIM offers unique material features that will increase melt temperatures, increase throughput, increase mixing, increase loading in the waste form, lower melter foot prints, eliminate melter corrosion and lower costs. These features not only enhance the technology for producing HLW forms, but also provide advantageous attributes to the waste form by allowing more durable alternatives to glass. It is concluded that glass ceramic waste forms that are tailored to immobilize fission products of HLW can be can be made from the HLW processed with the CCIM. The advantageous higher temperatures reached with the CCIM and unachievable with JHM allows the lanthanides, alkali, alkaline earths, and molybdenum to dissolve into a molten glass. Upon controlled cooling they go into targeted crystalline phases to form a glass ceramic waste form with higher waste loadings than achievable with borosilicate glass waste forms. Natural cooling proves to be too fast for the formation of all targeted crystalline phases.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2013

Off-gas adsorption model and simulation - OSPREY

A capability of accurately simulating the dynamic behavior of advanced fuel cycle separation processes is expected to provide substantial cost savings and many technical benefits. To support this capability, a modeling effort focused on the off-gas treatment system of a used nuclear fuel recycling facility is in progress. The off-gas separation consists of a series of scrubbers and adsorption beds to capture constituents of interest. Dynamic models are being developed to simulate each unit operation involved so each unit operation can be used as a stand-alone model and in series with multiple others. Currently, an adsorption model has been developed within Multi-physics Object Oriented Simulation Environment (MOOSE) developed at the Idaho National Laboratory (INL). Off-gas Separation and Recovery (OSPREY) models the adsorption of offgas constituents for dispersed plug flow in a packed bed under non-isothermal and non-isobaric conditions. Inputs to the model include gas composition, sorbent and column properties, equilibrium and kinetic data, and inlet conditions. The simulation outputs component concentrations along the column length as a function of time from which breakthrough data can be obtained. The breakthrough data can be used to determine bed capacity, which in turn can be used to size columns. In addition to concentration data, the model predicts temperature along the column length as a function of time and pressure drop along the column length. A description of the OSPREY model, results from krypton adsorption modeling and plans for modeling the behavior of iodine, xenon, and tritium will be discussed. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗