Survey of Dynamic Simulation Programs for Nuclear Fuel Reprocessing

The absence of any industrial scale nuclear fuel reprocessing in the U.S. has precluded the necessary driver for developing the advanced simulation capability now prevalent in so many other industries. Modeling programs to simulate the dynamic behavior of nuclear fuel separations and processing were originally developed to support the US government’s mission of weapons production and defense fuel recovery. Consequently there has been little effort is the US devoted towards improving this specific process simulation capability during the last two or three decades. More recent work has been focused on elucidating chemical thermodynamics and developing better models of predicting equilibrium in actinide solvent extraction systems. These equilibrium models have been used to augment flowsheet development and testing primarily at laboratory scales. The development of more robust and complete process models has not kept pace with the vast improvements in computational power and user interface and is significantly behind simulation capability in other chemical processing and separation fields

Beryllium Technology Research in the United States

While most active research involving beryllium in the United States remains tied strongly to biological effects, there are several areas of technology development in the last two years that should be mentioned. (1) Beryllium disposed of in soil vaults at the Idaho National Laboratory (INL) Radioactive Waste Management Complex (RWMC) has been encapsulated in-situ by high-temperature and pressure injection of a proprietary wax based material to inhibit corrosion. (2) A research program to develop a process for removing heavy metals and cobalt from irradiated beryllium using solvent extraction techniques has been initiated to remove components that prevent the beryllium from being disposed of as ordinary radioactive waste. (3) The JUPITER-II program at the INL Safety and Tritium Applied Research (STAR) facility has addressed the REDOX reaction of beryllium in molten Flibe (a mixture of LiF and BeF2) to control tritium, particularly in the form of HF, bred in the Flibe by reactions involving both beryllium and lithium. (4) Work has been performed at Los Alamos National Laboratory to produce beryllium high heat flux components by plasma spray deposition on macro-roughened substrates. Finally, (5) corrosion studies on buried beryllium samples at the RWMC have shown that the physical form of some of the corroded beryllium is very filamentary and asbestos-like. This form of beryllium may exacerbate the contraction of chronic beryllium disease

Synthesis and evaluation of an inorganic microsphere composite for the selective removal of {esc}p137{esc}scesium from acidic nuclear waste solutions /by Troy J. Tranter.

This work describes the results of a multi-year research and development effort to produce an inorganic ion exchange material for removing {esc}p137{esc}s Cs from acidic nuclear waste solutions. Various quantities of this waste exist throughout the United States, France, and Russia as a legacy of decades of nuclear fuel reprocessing and nuclear weapons production. For example, the Idaho National Laboratory (INL) currently stores about 900,000 gallons of acidic, high-level radioactive waste stemming from the various solvent decontamination processes associated with the reprocessing of naval reactor fuel assemblies. Internationally, the planned disposition path for these waste streams is solidification followed by storage in a geological repository. Since {esc}p137{esc}s Cs is a primary contributor to heat load and radiological dose, most treatment schemes involve removing this isotope from the bulk waste in order to facilitate handling and storage. However, since these liquid waste streams are highly acidic and ionic, they become problematic for any type of separation process. Consequently, an adsorbent or ion exchange material designed for use with these waste streams must be unique, having exceptional selectivity and stability in high radiation, temperature, and acid environments.;As result of this research effort, a new inorganic ion exchange composite consisting of ammonium molybdophosphate, (NH{esc}b4{esc}s){esc}b3{esc}sP(Mo{esc}b3{esc}sO{esc}b10{esc}s){esc}b4{esc}s*3H{esc}b2{esc}sO(AMP), synthesized within hollow aluminosilicate microspheres (AMP-C) has been produced. The selective cesium exchange capacity of this inorganic composite was evaluated using simulated sodium bearing waste solution as a surrogate for the tank waste currently stored at the INL. Equilibrium isotherms obtained from these experiments were very favorable for cesium uptake and indicated maximum cesium loading of approximately 9 % by weight of dry AMP. Column tests were performed using bench-scale columns and complete breakthrough curves were obtained from these tests. The dynamic capacity of the columns was determined to be approximately 2.5 g Cs/kg exchanger (18.8 millimole/kg) for the feed concentrations of interest.Thesis (Ph. D., Chemical Engineering)--University of Idaho, May 2006

System Design Description and Requirements for Modeling the Off-Gas Systems for Fuel Recycling Facilities

This document provides descriptions of the off-gases evolved during spent nuclear fuel processing and the systems used to capture the gases of concern. Two reprocessing techniques are discussed, namely aqueous separations and electrochemical (pyrochemical) processing. The unit operations associated with each process are described in enough detail so that computer models to mimic their behavior can be developed. The document also lists the general requirements for the desired computer models