A Superabsorbing Hydrogel for Radiological Dispersal Device ("Dirty Bomb") Cleanup. NADIA VASQUEZ (Richard J. Daley College, Chicago, IL 60652) DR. MICHAEL KAMINSKI (Argonne National Laboratory, Argonne, IL 60439)

Radiological decontamination technologies are needed for non-destructive removal of radioactivity from porous surfaces such as concrete and marble. We are configuring a novel process for post-restoration of an RDD from porous materials in the event of a terrorist attack. The optimized process would involve three steps: (1) remove surface bound species and penetrate the pore structure to free radionuclide ions from the surface and into the pore water; (2) pull water from the pore structure with a superabsorbing hydrogel, and; (3) remove the radioactivity-loaded gel by wet vacuum. We studied decontamination properties in designing an optimal polymer gel formulation. We report performance parameters of polymer candidates for aqueous solution absorbency with a gravimetric analysis of swelling capacity of gel formulations ("tea bag test"). We found the polymer water absorbency is dependent on the effect of ions or chelators in the following wash solutions: NH4Cl, CaCl2, and deionized H2O. We observed increased absorption capacity using smaller grain sized polymer than commercially distributed polymer. The resulting superabsorbent retention capacity calculations suggest that the critical absorption time of polymer formulations is within ten minutes of immersion. The absorbency of polymer candidates is affected by cross-linked and linear formulation ratios. Furthermore, double immersion did not affect subsequent retention of the polymer.

Algorithm Development for the Image Reconstruction of Tomography Data from Neutron Radiography. SUSAN KING (University of Tennessee, Knoxville, TN 37916) JEFF SANDERS (Argonne National Laboratory, Argonne, IL 60439)

Neutron radiography is a unique method of nondestructively examining an object. Neutrons are suited to examine materials that other techniques, such as x-rays, cannot examine well. Like x-rays, they can give a very useful image of what the structure of an object looks like, but how neutrons and x-rays interact with matter is very different. Neutrons are attenuated preferentially by light materials, such as hydrogen and boron, while x-rays are attenuated by heavy, dense materials. As a result, neutrons are more suited to radiography of higher atomic number elements and in many cases, high radiation fields. This paper will examine the methods used to reconstruct tomography data gathered from neutron radiography, as well as explain how that data is collected. It will compare tomography reconstruction methods using standard phantoms and data used in x-ray radiography. Though the algorithms developed are intended to be used for neutron radiography in NRAD, the TRIGA neutron radiography reactor at Argonne National Laboratory West, the well defined x-ray data and phantoms will be used to perform initial evaluations of the reconstruction methods. Finally, suggestions on further improvements of the algorithms and descriptions of further work will be presented.

Creation of Theoretical Phase Diagrams to Model New Nuclear Fuel Alloys. KRISTIN BRINEY (DePauw University, Greencastle, IN 46135) DR J RORY KENNEDY (Argonne National Laboratory, Argonne, IL 60439)

The software Thermo-Calc is analyzed for its effectiveness in creating phase diagrams of transuranic fuel alloys. The software is found to be very powerful and flexible in creating diagrams, though rather database dependent. The lack of information on transuranic elements in the databases results in poor diagrams that do not match experimentally determined values. On a very basic level, however, data for single elements is shown to be accurate. The conclusion is made that a better, more thorough database is required to create useful diagrams. Also, some experimental data will need to be added to the program in order to verify the accuracy of the resulting diagrams.

Flow Testing for Reduced Enrichment Research and Test Reactors (RERTR) Fuel. ADAM ROBINSON (Oregon State University, Corvallis, OR 97330) DAN WACHS (Argonne National Laboratory, Argonne, IL 60439)

The Reduced Enrichment Research and Test Reactor program was created to take the highly enriched uranium (HEU) fuel that operates many of today’s research and test reactors and reduce them to low enriched uranium (LEU) fuels. Argonne National Lab has been a major part of this effort. They have designed and fabricated two types of fuels that they hope will satisfy this goal and plan on doing radiation testing on these fuels in the Advanced Test Reactor at Idaho National Engineering Laboratory. In order to model this new fuels behavior in the reactor, accurate thermal modeling must done first. In order to perform this thermal modeling, flow data had to be obtained for these fuels. A model of the fuel assembly was created and a system was built to test the flow over this fuel assembly as a function of differential core pressure, flow rate, and flow restriction due to orifices on the fuel assembly channel. This was also built at Argonne National Lab and was operated at the Idaho State University Thermal Hydraulics Laboratory. An array of orifice sizes were tested and loss coefficients were calculated for the different tested systems. With the obtained flow values and loss coefficients, accurate thermal models can now be created and analyzed before the fuel is sent for irradiation.

Modeling Salt/Zeolite Ion Exchange. JILL RYDALCH (Idaho State University, Pocatello, ID 83201) MICHAEL SIMPSON (Argonne National Laboratory, Argonne, IL 60439)

Ion exchange between molten chloride salt and zeolite-A is being developed as a possibility for the treatment of spent nuclear fuel. A two-site model for monovalent salt/zeolite-A ion exchange has been developed by Simpson and Gougar and a program written in matlab to test it against the experimental data. The results have been compared with the results of Simpson’s excel spreadsheet. Parameters have been found that provide the best fit for the data. The solutions of the monovalent model have been compared with Gougar’s multivalent model and the results plotted. Both Simpson’s monovalent and Gougar’s multivalent models adequately and comparably fit the monovalent data. In order to compare the two multivalent equations Simpson’s multivalent model needs to put into a matlab code and the results generated from there.

The Development of Finite Element Based Models to Predict the Performance of Proposed Generation IV Gas-Cooled Fast Reactor Fuels. MARY ERNESTI (University of Missouri - Rolla, Rolla, MO 65401) DAN WACHS (Argonne National Laboratory, Argonne, IL 60439)

The thermal and mechanical performance of Gas-Cooled Fast Reactor (GFR) dispersion fuel forms for the Gen IV GFR reactor were evaluated using Finite Element Analysis (FEA) tools. FEA models of a 1/6 segment of a hexagonal element containing either hexagonal or cylindrical coolant flow channels were drawn in a 3-D modeling program. The model fuel elements were assumed to be composed of a SiC matrix and used to complete steady-state thermal and stress analysis. Due to the ceramic properties of SiC, tensile stress was considered the limiting factor in the analysis of the different geometries. A chosen maximum principal stress of 150 MPa was set as the material limit, with a safety factor of three. A goal of 50 MPa was set. After reviewing the model data from the numerous finite element models analyses, the cylindrical geometry was chosen as the more desirable design. Parametric studies on the density of coolant channels and web thickness were then performed on the cylindrical coolant flow models. As the density of the coolant channels increased, the maximum principal stress decreased. A full scale hexagonal core assembly with several different side lengths was also modeled. Future work will focus on more specific parametric studies of the cylindrical flow channel geometry concerning thermal conductivity, convection coefficients, and the thickness of the fuel elements.