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Purpose: The purpose of this study was to establish a method to perform multidimensional radiochromic film measurements of (106)Ru plaques and to benchmark the resulting dose distributions against Monte Carlo simulations (MC), microdiamond, and diode measurements., Methods: Absolute dose rates and relative dose distributions in multiple planes were determined for three different plaque models (CCB, CCA, and COB), and three different plaques per model, using EBT3 films in an in-house developed polystyrene phantom and the mcnp6 MC code. Dose difference maps were generated to analyze interplaque variations for a specific type, and for comparing measurements against MC simulations. Furthermore, dose distributions were validated against values specified by the manufacturer (BEBIG) and microdiamond and diode measurements in a water scanning phantom. Radial profiles were assessed and used to estimate dosimetric margins for a given combination of representative tumor geometry and plaque size., Results: Absolute dose rates at a reference depth of 2 mm on the central axis of the plaque show an agreement better than 5% (10%) when comparing film measurements (mcnp6) to the manufacturer's data. The reproducibility of depth-dose profile measurements was <7% (2 SD) for all investigated detectors and plaque types. Dose difference maps revealed minor interplaque deviations for a specific plaque type due to inhomogeneities of the active layer. The evaluation of dosimetric margins showed that for a majority of the investigated cases, the tumor was not completely covered by the 100% isodose prescribed to the tumor apex if the difference between geometrical plaque size and tumor base ≤4 mm., Conclusions: EBT3 film dosimetry in an in-house developed phantom was successfully used to characterize the dosimetric properties of different (106)Ru plaque models. The film measurements were validated against MC calculations and other experimental methods and showed a good agreement with data from BEBIG well within published tolerances. The dosimetric information as well as interplaque comparison can be used for comprehensive quality assurance and for considerations in the treatment planning of ophthalmic brachytherapy.

Purpose: The response of alanine solid state dosimeters to ionizing radiation strongly depends on particle type and energy. Due to nuclear interactions, neutron fields usually also consist of secondary particles such as photons and protons of diverse energies. Various experiments have been carried out in three different neutron beams to explore the alanine dose response behavior and to validate model predictions. Additionally, application in medical neutron fields for boron neutron capture therapy is discussed., Methods: Alanine detectors have been irradiated in the thermal neutron field of the research reactor TRIGA Mainz, Germany, in five experimental conditions, generating different secondary particle spectra. Further irradiations have been made in the epithermal neutron beams at the research reactors FiR 1 in Helsinki, Finland, and Tsing Hua open pool reactor in HsinChu, Taiwan ROC. Readout has been performed with electron spin resonance spectrometry with reference to an absorbed dose standard in a (60)Co gamma ray beam. Absorbed doses and dose components have been calculated using the Monte Carlo codes fluka and mcnp. The relative effectiveness (RE), linking absorbed dose and detector response, has been calculated using the Hansen & Olsen alanine response model., Results: The measured dose response of the alanine detector in the different experiments has been evaluated and compared to model predictions. Therefore, a relative effectiveness has been calculated for each dose component, accounting for its dependence on particle type and energy. Agreement within 5% between model and measurement has been achieved for most irradiated detectors. Significant differences have been observed in response behavior between thermal and epithermal neutron fields, especially regarding dose composition and depth dose curves. The calculated dose components could be verified with the experimental results in the different primary and secondary particle fields., Conclusions: The alanine detector can be used without difficulty in neutron fields. The response has been understood with the model used which includes the relative effectiveness. Results and the corresponding discussion lead to the conclusion that application in neutron fields for medical purpose is limited by its sensitivity but that it is a useful tool as supplement to other detectors and verification of neutron source descriptions.

Purpose: In order to build up a reliable dose monitoring system for boron neutron capture therapy (BNCT) applications at the TRIGA reactor in Mainz, a computer model for the entire reactor was established, simulating the radiation field by means of the Monte Carlo method. The impact of different source definition techniques was compared and the model was validated by experimental fluence and dose determinations., Methods: The depletion calculation code origen2 was used to compute the burn-up and relevant material composition of each burned fuel element from the day of first reactor operation to its current core. The material composition of the current core was used in a mcnp5 model of the initial core developed earlier. To perform calculations for the region outside the reactor core, the model was expanded to include the thermal column and compared with the previously established attila model. Subsequently, the computational model is simplified in order to reduce the calculation time. Both simulation models are validated by experiments with different setups using alanine dosimetry and gold activation measurements with two different types of phantoms., Results: The mcnp5 simulated neutron spectrum and source strength are found to be in good agreement with the previous attila model whereas the photon production is much lower. Both mcnp5 simulation models predict all experimental dose values with an accuracy of about 5%. The simulations reveal that a Teflon environment favorably reduces the gamma dose component as compared to a polymethyl methacrylate phantom., Conclusions: A computer model for BNCT dosimetry was established, allowing the prediction of dosimetric quantities without further calibration and within a reasonable computation time for clinical applications. The good agreement between the mcnp5 simulations and experiments demonstrates that the attila model overestimates the gamma dose contribution. The detailed model can be used for the planning of structural modifications in the thermal column irradiation channel or the use of different irradiation sites than the thermal column, e.g., the beam tubes.

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.