Monte Carlo simulation applied to uncertainties in iodine-123 assay and thyroid uptake measurement

The Monte Carlo method can be used to simulate deterministic processes such as the interaction of nuclear particles with matter, using a probabilistic approach. For radiation transport, the method is realised using Monte Carlo radiation transport codes. Performing simulations often has a distinct advantage over experimental studies, especially when physical measurements would be impracticable or even impossible to undertake. The application of Monte Carlo methods to nuclear medicine problems can therefore be seen as advantageous. 123I is a commonly utilised radionuclide for nuclear medicine diagnostic imaging and non-imaging investigations. The radioactive decay scheme of 123I presents particular challenges in terms of accurately determining the activity of an 123I based radiopharmaceutical prior to administration. The aims of this project were to use Monte Carlo modelling techniques to investigate the uncertainties associated with 123I assay and thyroid uptake assessment using this radionuclide. A model of the UK secondary standard calibrator was created using MCNP5 code and validated against the physical calibrator. A sensitivity curve for the instrument was created through simulation and the effect of measuring 123I using a vial and a syringe investigated. Up to 21% variation was seen between the vial and the syringe geometries studied. A thyroid uptake counter model was produced to determine uncertainties in thyroid uptake assessment for a number of different parameters. Variations in collimator to Page | iv neck distance, horizontal displacement of the detector and increasing depth of the thyroid gland in tissue were shown to affect the accuracy of uptake measurement. A depth-based correction for the thyroid gland was derived from the differential counts in the x-ray and gamma peaks of 123I. Such an approach could be utilised in clinical practice to correct for the depth of the thyroid gland in the neck

Non-vascular interventional procedures: effective dose to patient and equivalent dose to abdominal organs by means of dicom images and Monte Carlo simulation

This study evaluates X-ray exposure in patient undergoing abdominal extra-vascular interventional procedures by means of Digital Imaging and COmmunications in Medicine (DICOM) image headers and Monte Carlo simulation. The main aim was to assess the effective and equivalent doses, under the hypothesis of their correlation with the dose area product (DAP) measured during each examination. This allows to collect dosimetric information about each patient and to evaluate associated risks without resorting to in vivo dosimetry. The dose calculation was performed in 79 procedures through the Monte Carlo simulator PCXMC (A PC-based Monte Carlo program for calculating patient doses in medical X-ray examinations), by using the real geometrical and dosimetric irradiation conditions, automatically extracted from DICOM headers. The DAP measurements were also validated by using thermoluminescent dosimeters on an anthropomorphic phantom. The expected linear correlation between effective doses and DAP was confirmed with an R(2) of 0.974. Moreover, in order to easily calculate patient doses, conversion coefficients that relate equivalent doses to measurable quantities, such as DAP, were obtained

Validation and study of different parameters in the simulation of diagnostic X-ray spectra using the MCNPX code

In radiology, knowing the X-ray spectrum characteristics makes it possible to estimate the absorbed dose in the patient and to improve image quality. In this study, an X-ray generator was proposed using the MCNPX code and to validate it, the simulated spectrum was compared to the data provided from AAPM Task Group 195, which resulted in a percentage difference of 8.7%. Furthermore, several X-ray spectra were generated and compared to the spectra obtained from commercially available softwares as xpecgen and SpekCalc. The percentage differences were of the order of 13% in comparison with SpekCalc and 8% with xpecgen. The major differences obtained between those spectra were concentrated in the region of characteristic peaks, independently if variations in electron beam energy, target angle or filtration thickness were performed

A Flag-Based Algorithm for Explosives Detection in Sea-Land Cargo Containers using Active Neutron Interrogation.

The high volume and minimal screening of sea land cargo containers presents a vulnerability in which explosive devices may be smuggled across national borders. Fast neutrons are a strong candidate for use in container screening due to their high target penetration and ability to discriminate between materials of low atomic mass, such as explosives and non metallic container contents. An algorithm has been developed that uses flags, calculated from specific measurements of the reflected neutrons and photons produced during active neutron interrogation, to discern explosives hidden in cargo containers. Steps in algorithm development included Monte Carlo simulations for scatter characterization, identification of flags in idealized scenarios, refinement of flags in realistic scenarios, combining the flags into a detection algorithm, and evaluation of the algorithm and associated detection system. Simulations compared favorably with small scale neutron scatter measurements using the explosives surrogate, melamine. The detection algorithm included corrections for different types of cargo contents and cargo inhomogeneity, surrounding environment, and realistic neutron sources and radiation detectors. The proposed algorithm has two variations, one of which can be easily implemented with today’s technology. The proposed scanning system utilizes a shielded 14.1 MeV neutron generator, eleven large liquid scintillators neutron detectors, and several inorganic scintillators for photon spectroscopy. This system should cost less than $1M to install and dose estimates fall well within acceptable levels for both operators and smuggled persons. Algorithm performance has been quantified with various explosive sizes and positions, as well as heterogeneous cargo configurations, with typical minimum detectable amounts not exceeding 200 kg.Ph.D.Nuclear Engineering & Radiological SciencesUniversity of Michigan, Horace H. Rackham School of Graduate Studieshttp://deepblue.lib.umich.edu/bitstream/2027.42/91602/1/alehnert_1.pd

Oramed: optimization of radiation protection of medical staff

Postprint (published version

Advanced brachytherapy dosimetric considerations

Thesis (Ph. D.)--Harvard-MIT Division of Health Sciences and Technology, 2008.Includes bibliographical references (p. 131-139).The practice of brachytherapy and brachytherapy dosimetry was investigated with emphasis on evaluations of dose distributions and shielding considerations for both photon- and neutron-emitting radionuclides. Monte Carlo simulation methods were employed to calculate dose distributions for virtual and commercial brachytherapy sources. Radionuclides studied were 103Pd, 1251, 131Cs, 137Cs, 169b, 192Ir, and 252Cf. 252Cf sources also emit neutrons from spontaneous fission. The brachytherapy dosimetry protocol recommended by the American Association of Physicists in Medicine was followed and evaluated for conditions of partial scatter (non-infinite media) and material inhomogeneities, both commonly encountered in brachytherapy treatment. Furthermore, energy-dependent characteristics of dosimetry parameters were evaluated and reference calculations performed for virtual photon and neutron sources. These findings were applied to three clinical brachytherapy cases: eye plaques using 103Pd, 125I, and 131Cs; high-dose rate 252Cf treatment; and, 2 Cf plaques for superficial lesions. For eye plaques, material heterogeneities were significant for each radionuclide with dose reduction at 5 mm of 18%, 11%, and 10% for P03pd, 125I, and 131Cs, respectively. For a proposed highdose rate 252Cf source (5mm length), relative brachytherapy dosimetry parameters were found to be similar to those obtained for a low-dose rate Applicator Tube-type source (15 mm length). Considering 252Cf plaque brachytherapy when partial scatter conditions were accounted for, central axis equivalent dose rate decreased by 11 ± 1% and 7 ± 2% for depths of 4 to 50 mm, respectively.(cont.) The ratio of neutron dose to total physical dose was 70 ± 1% and 57 ± 2% for depths of 4 and 50 mm, respectively, while the fractional dose-equivalent due to neutrons was 93 + 1% and 89 ± 2% at these depths, respectively. Finally, shielding requirements for a clinical high-dose rate 252Cf source were explored for common shielding materials and a linear accelerator vault. Lead, polyethylene, and borated polyethylene were evaluated for neutron, primary photon, and secondary photon attenuation. Half-value layers of 0.70, 0.15, and 0.13 m were obtained for lead, polyethylene, and borated polyethylene, respectively. A linear accelerator vault was found to adequately shield up to a 5 mg 252Cf source for regular clinical use.by Christopher S. Melhus.Ph.D