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Discrete beta dose kernel matrices for nuclides applied in targeted radionuclide therapy (TRT) calculated with MCNP5.

PURPOSE
Radiopharmaceuticals administered in targeted radionuclide therapy (TRT) rely to a great extent not only on beta-emitting nuclides but also on emitters of monoenergetic electrons. Recent advances like combined PET/CT devices, the consequential coregistration of both data, the concept of using beta couples for diagnosis and therapy, respectively, as well as the development of voxel models offer a great potential for developing TRT dose calculation systems similar to those available for external beam treatment planning. The deterministic algorithms in question for this task are based on the convolution of three-dimensional matrices, one representing the activity distribution and the other the dose point kernel. This study aims to report on three-dimensional kernel matrices for various nuclides used in TRT.


METHODS
The Monte Carlo code MCNP5 was used to calculate discrete dose kernels of beta particles including the contributions from their respective secondary radiation in soft tissue for the following nuclides: 32P, 33P, 67Cu, 89Sr, 90Y, 103Rh9m), 131I, 177Lu, 186Re, and 188Re. For each nuclide a kernel cube of 10 x 10 x 10 mm3 was calculated, the dimensions of a voxel being 1 mm3. Additional kernels with voxel sizes of 3 x 3 x 3 mm3 were simulated.


RESULTS
Comparison with the S-value data regarding 32P, 89Sr, 90Y, and 131I of the MIRD committee which were calculated with the EGS4 code showed a very good agreement, the secondary particle transport of 90Y being the only exception. Documented analytical kernels on the other side show deviations very close and very far to the source.


CONCLUSIONS
The good accordance with the only discrete dose kernels published up to date justifies the method chosen. Together with the additional six nuclides, this report provides a considerable database for three-dimensional kernel matrices with regard to beta radionuclides applied in TRT. In contrast to analytical dose point kernels, the discrete kernels elude the problem of overestimation near the source and take energy depositions into account, which occur beyond the range of the continuous-slowing-down approximation (csda range). Recalculation of the 1 x 1 x 1 mm3 kernels to other dose kernels with varying voxel dimensions, cubic or noncubic, is shown to be easily manageable and thereby provides a resolution-independent system of dose calculation.