ABSTRACT
Monte Carlo simulations are increasingly used to reconstruct dose distributions in radiotherapy research studies. Many studies have used the MCNPX Monte Carlo code with a mesh tally for dose reconstructions. However, when the number of voxels in the simulated patient anatomy is large, the computation time for a mesh tally can become prohibitively long. The purpose of this work was to test the feasibility of using lattice tally instead of mesh tally for whole-body dose reconstructions. We did this by comparing the dosimetric accuracy and computation time of lattice tallies with those of mesh tallies for craniospinal proton irradiation. The two tally methods generated nearly identical dosimetric results, within 1% in dose and within 1 mm distance-to-agreement for 99% of the voxels. For a typical craniospinal proton treatment field, simulation speed was 4 to 17 times faster using the lattice tally than using the mesh tally, depending on the numbers of proton histories and voxels. We conclude that the lattice tally is an acceptable substitute for the mesh tally in dose reconstruction, making it a suitable potential candidate for clinical treatment planning.
ABSTRACT
Using Dawn's Gamma Ray and Neutron Detector, we tested models of Vesta's evolution based on studies of howardite, eucrite, and diogenite (HED) meteorites. Global Fe/O and Fe/Si ratios are consistent with HED compositions. Neutron measurements confirm that a thick, diogenitic lower crust is exposed in the Rheasilvia basin, which is consistent with global magmatic differentiation. Vesta's regolith contains substantial amounts of hydrogen. The highest hydrogen concentrations coincide with older, low-albedo regions near the equator, where water ice is unstable. The young, Rheasilvia basin contains the lowest concentrations. These observations are consistent with gradual accumulation of hydrogen by infall of carbonaceous chondrites--observed as clasts in some howardites--and subsequent removal or burial of this material by large impacts.
ABSTRACT
A new Monte Carlo method is being developed to calculate eigenfunction fluxes in critical or near-critical nuclear systems. The correct estimation of fluxes is essential for radiation protection and shielding near these systems, in addition to isotope production, isotope depletion, nuclear criticality and other applications. The proposed method applies to Monte Carlo criticality eigenvalue calculations in which the fission sites in one generation are used as fission sources in subsequent generations. The usual Monte Carlo power iteration method for such problems often calculates fluxes (eigenfunctions) that are inaccurate and very different in symmetric parts of a problem geometry. The proposed method calculates flux distributions by estimating an approximate fission matrix. The way the fission matrix is estimated and used differs from other recent works. Preliminary results are promising.