Biomedical subjects
Publications and source records attributed to E E Furhang.
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New source designs of encapsulated low-energy gamma emitting nuclides for permanent implants require dosimetric analysis and calibration standardization. The dosimetry measurements can be incorporated into a treatment planning system by fitting the data. The use of a fitting function whose behavior at range limits mimics the physical phenomena, using as few parameters as possible, eliminates noisy outliers and lends credence to calculations beyond the measured range. Clinical implementation of the new sources also requires benchmarking against existing sources, where the current clinical experience lies. We present an analysis of measured dosimetry data for three brachytherapy sources recently available from North American Scientific, Inc. (North Hollywood, CA): 103Pd source model MED3633 ("PdGold"), 125I source models MED3631-A/M ("IoGold-AM") and MED3631-A/S ("IoGold-AS"). Using the formalism of the Interstitial Collaborative Working Group (ICWG) the radial dose function, g(r), the anisotropy function, F(r, theta), and the anisotropy factor, phi an(r), were previously evaluated from measurements of each source design. In this report we use fitting functions whose forms are chosen to approach reasonable values at data limits. These forms are quite similar to those used in a previous analysis of TG43 Iodine and palladium compendium data. Fitting parameter results for each function are provided for each brachytherapy source model. Fit-data discrepancies are smaller than measurement uncertainties, meaning that incorporation into treatment planning systems will not introduce significant errors in clinical use. Current clinical experience is based on the Theragenics (Norcross, GA) 103Pd seed ("PdThera"), and the Nycomed-Amersham (Arlington Heights, IL) 125I seed models 6711 ("6711") and 6702 ("6702"). The new sources are benchmarked against these seeds.
This work was undertaken to expedite implementation of the AAPM Task Group 43 recommendations, which call for significant modifications in the way dose is calculated for interstitial sources of 192Ir, 125I, and 103Pd as well as significant changes in the dose rate constant for 125I sources. The TG43 recommendations include a new formalism for dose calculation at points defined by the radial distance, r, from the source center and the angle, theta, that such a radius makes with the source axis. For each source type, values are tabulated for the radial dose function, the anisotropy function, and the anisotropy factor. The TG43 report includes fitting functions for the radial dose function in the form of polynomials, which are poorly behaved outside the range of fitted data. No functions are offered for the anisotropy function data or the anisotropy factor data, both of which could profit from some smoothing by such functions. We have found a double exponential fit to the radial dose function that not only approximates the data adequately but also appropriately approaches zero for very large distances. The anisotropy function is conveniently fit with a form of type 1 - f(r,theta)cos(theta)e(cr), which is exactly 1 at theta=90 degrees and approaches 1 for large r (for c<0), where f(r,theta) is a selected polynomial in the two variables. The form chosen for the anisotropy factor was 1 - (a+br)e(cr), which appropriately approaches 1 for large r (and c<0). Functional fits of these types are expected to facilitate implementation of TG43 recommendations, in that they may be either incorporated into dose algorithms or used to generate lookup tables of either the x, y or the r, theta format.
UNLABELLED: Since 1962, Memorial Sloan Kettering Cancer Center has used an individually optimized dosimetry method for patients with thyroid carcinoma undergoing radioiodine therapy. This traditional dosimetry method involves a determination of the maximum tolerated activity or the activity that will deliver 2 Gy to the blood (A(max)), and the corresponding ablative lesion dose (D(lesion)). However, the traditional calculations of A(max) and D(lesion) were based on empirical assumptions. The objective of this work was to develop a dosimetry method that eliminates these assumptions by incorporating patient kinetics and that is not restricted to 131I as a tracer and therapeutic agent. METHODS: Patient kinetics were incorporated into the dosimetry algorithm by fitting parameters to patient clearance measurements. The radioiodines 123I, 124I, 125I and 131I were accommodated as tracers and therapeutic agents by incorporating their physical half lives and by precalculating photon-absorbed fractions for these radionuclides for several thousand patient geometries using Monte Carlo simulations. RESULTS: A(max) and D(lesion) have been calculated using the traditional and new method for a group of patients, and errors associated with each of the above assumptions were examined. Assuming that the initial blood activity is distributed instantaneously in 5 L was found to introduce an error in A(max) of up to 30%, whereas assuming physical decay beyond the last data point introduced an error of up to 50%. CONCLUSION: Individualized fitting of clearance data is a practical method to accurately account for inter-patient kinetics variations. The substitution of standard kinetics beyond measured data might lead to substantial errors in estimating A(max) and D(lesion). In addition, gamma camera images, rather than neck probe readings, should be used to determine lesion uptakes for thyroid cancer patients.
In internal emitter therapy, an accurate description of the absorbed dose distribution is necessary to establish an administered dose-response relationship, as well as to avoid critical organ toxicity. This work describes the implementation of a dosimetry method that accounts for the radionuclide decay spectrum, and patient-specific activity and density distributions. The dosimetry algorithm is based on a Monte Carlo procedure that simulates photon and electron transport and scores energy depositions within the patient. The necessary input information may be obtained from a registered set of CT and SPECT or PET images. The algorithm provides the absorbed dose rate for the radioactivity distribution provided by the SPECT or PET image. The algorithm was benchmarked by reproducing dosimetric quantities using the Medical Internal Radionuclide Dose (MIRD) Committee's Standard Man phantom and was used to calculate absorbed dose distributions for representative case studies.
Photon point dose kernels and absorbed fractions were generated in water for the full photon emission spectrum of each radionuclide of interest in nuclear medicine, by simulating the transport of particles using Monte Carlo. The kernels were then fitted to a mathematical expression. Absorbed fractions for point sources were obtained by integrating the kernels over spheres. Photon dose kernels and absorbed fractions were generated for the following radionuclides: I-123, I-124, I-125, I-131, In-111, Cu-64, Cu-67, Ga-67, Ga-68, Re-186, Re-188, Sm-153, Sn-117m, Tc-99m. The Monte Carlo simulation was verified by comparing the dose kernels to published monoenergetic photon kernels. Further validation was obtained by generating an I-125 brachytherapy seed kernel and comparing it with published data. Since Monte Carlo simulation was initialized by sampling from the complete photon spectra of these radionuclides, interpolation between monoenergetic kernels and absorbed fractions was not required. The absorbed-fraction due to uniform spherical distributions can be directly applied for use in internal dosimetry. In addition, the kernels can be used as input for three-dimensional internal dosimetry calculations.
In internal emitter therapy, an accurate description of the absorbed dose distribution is necessary to establish an administered dose-response relationship, as well as to avoid critical organ toxicity. Given a spatial distribution of cumulated activity, an absorbed dose distribution that accounts for the effects of attenuation and scatter can be obtained using a Monte Carlo method that simulates particle transport across the various densities and atomic numbers encountered in the human body. Patient-specific information can be obtained from CT and SPECT or PET imaging. Since the data from these imaging modalities is discrete, it is necessary to develop a technique to efficiently transport particles across discrete media. The Monte Carlo-based algorithm presented in this article produces accurate absorbed dose distributions due to patient-specific density and radionuclide activity distributions. The method was verified by creating CT and SPECT arrays for the Medical Internal Radionuclide Dose (MIRD) Committee's Standard Man phantom, and reproducing the spatially averaged specific absorbed fractions reported in MIRD Pamphlet 5. The algorithm was used to investigate the implications of replacing a mean absorbed dose with a distribution, and of neglecting atomic number and density variations for various patient geometries and energies. For example, the I-131 specific absorbed fraction for spleen to liver is the same as for liver to spleen, yet the distributions were different. Furthermore, neglecting atomic number variations across the vertebral bone led to an overestimation of I-125 absorbed dose by an order of magnitude, while no error was observed for I-131.
Mean mass energy absorption coefficient ratios of acrylic, polystyrene, and water to air, were calculated using Monte Carlo generated energy spectra. The energy spectra were calculated for 4- to 50-MV x-ray beams, from machines using flattening filters and scanning beams. The validity of these spectra was verified by comparing the measured ionization ratios with the calculated values. The agreement was found to be within 1.9%. For beams of energy below 6 MV, our estimates of the mean mass energy absorption coefficient ratios agree well with those recommended by the TG-21 protocol. For higher energy beams, the discrepancy increases to about 3%. It was found that the discrepancy is attributable to the different spectra used in these calculations.