Researchers at the Lawrence Berkeley National Laboratory have invented ARUBA (Arsenic Removal Using Bottom Ash) a material that effectively and affordably removes high concentrations of arsenic from contaminated groundwater. The technology is cost-effective because the substrate?bottom ash from coal fired power plants?is a waste material readily available in South Asia. During fieldwork in four sub-districts of Bangladesh, ARUBA reduced groundwater arsenic concentrations as high as 680 ppb to below the Bangladesh standard of 50 ppb. Key results from three trips in Bangladesh and one trip to Cambodia include (1) ARUBA removes more than half of the arsenic from contaminated water within the first five minutes of contact, and continues removing arsenic for 2-3 days; (2) ARUBA?s arsenic removal efficiency can be improved through fractionated dosing (adding a given amount of ARUBA in fractions versus all at once); (3) allowing water to first stand for two to three days followed by treatment with ARUBA produced final arsenic concentrations ten times lower than treating water directly out of the well; and (4) the amount of arsenic removed per gram of ARUBA is linearly related to the initial arsenic concentration of the water. Through analysis of existing studies, observations, and informal interviews in Bangladesh, eight design strategies have been developed and used in the design of a low-cost, community-scale water treatment system that uses ARUBA to remove arsenic from drinking water. We have constructed, tested, and analyzed a scale version of the system. Experiments have shown that the system is capable of reducing high levels of arsenic (nearly 600 ppb) to below 50 ppb, while remaining affordable to people living on less than $2 per day. The system could be sustainably implemented as a public-private partnership in rural Bangladesh.

The current state of the art in reactor physics methods to assess safety, fuel failure, and operability margins for Design Basis Accidents (DBAs) for Light Water Reactors (LWRs) rely upon the coupling of nodal neutronics and one-dimensional thermal hydraulic system codes. The neutronic calculations use a multi-step approach in which the assembly homogenized macroscopic cross sections and kinetic parameters are first calculated using a lattice code for the range of conditions (temperatures, burnup, control rod position, etc...) anticipated during the transient. The core calculation is then performed using the few group cross sections in a core simulator which uses some type of coarse mesh nodal method. The multi-step approach was identified as inadequate for several applications such as the design of MOX cores and other highly hetereogeneous, high leakage core designs. Because of the considerable advances in computing power over the last several years, there has been interest in high-fidelity solutions of the Boltzmann Transport equation. A practical approach developed for high-fidelity solutions of the 3D transport equation is the 2D-1D methodology in which the method of characteristics (MOC) is applied to the heterogeneous 2D planar problem and a lower order solution is applied to the axial problem which is, generally, more uniform. This approach was implemented in the DeCART code. Recently, there has been interest in extending such approach to the simulations of design basis accidents, such as control rod ejection accidents also known as reactivity initiated accidents (RIA). The current 2D-1D algorithm available in DeCART only provide 1D axial solution based on the diffusion theory whose accuracy deteriorates in case of strong flux gradient that can potentially be observed during RIA simulations.

The primary ojective of the dissertation is to improve the accuracy and range of applicability of the DeCART code and to investigate its ability to perform a full core transient analysis of a realistic RIA.

The specific research accomplishments of this work include:

* The addition of more accurate 2D-1D coupling and transverse leakage splitting options to avoid the occurrence of negative source terms in the 2D MOC equations and the subsequent failure of the DeCART calculation and the improvement of the convergence of the 2D-1D method.

* The implementation of a higher order transport axial solver based on NEM-Sn derivation of the Boltzmann equation.

* Improved handling of thermal hydraulic feedbacks by DeCART during transient calculations.

* A consistent comparison of the DeCART transient methodology with the current multistep approach (PARCS) for a realistic full core RIA.

An efficient direct whole core transport calculation method involving the NEM-Sn formulation for the axial solution and the MOC for the 2-D radial solution was developed. In this solution method, the Sn neutron transport equations were developed within the framework of the Nodal Expansion Method. A RIA analysis was performed and the DeCART results were compared to the current generation of LWR core analysis methods represented by the PARCS code. In general there is good overall agreement in terms of global information from DeCART and PARCS for the RIA considered. However, the higher fidelity solution in DeCART provides a better spatial resolution that is expected to improve the accuracy of fuel performance calculations and to enable reducing the margin in several important reactor safety analysis events such as the RIA.