Nuclear Medicine Dosimetry Research Articles

TH-AB-BRA-10: Clinical Implementation of a Grid-Based Boltzmann Solver with Adaptive Meshing for Nuclear Medicine Dosimetry

Purpose: We previously compared the Grid-Based Boltzmann Solver (GBBS) with DOSXYZnrc Monte Carlo (MC) for beta and gamma sources in uniform materials and along material interfaces. The current standard in nuclear medicine is MIRD, which only calculates mean organ doses assuming uniform uptake in anthropomorphic phantoms. This work extends GBBS to clinical nuclear medicine for fast and accurate patient-specific voxel-level dosimetry. Methods: We compared the GBBS with MC using a patient’s post-therapy quantitative 90Y microsphere bremsstrahlung SPECT/CT.The GBBS (Attila™ v8.0.0) requires a tetrahedral mesh as input instead of voxels. Adaptive tetrahedral meshes were generated with TetGen v1.5.0 by folding CT and SPECT images into a mesh sizing function; the function was defined at each node on a background mesh with nodes defined at each SPECT/CT voxel. Target tetrahedral edge lengths varied from 0.5 to 8 cm and were adapted on SPECT activity level, CT material, and material gradient.GBBS used 30 electron energy groups, 32 angles, and 67,000 tetrahedrons. GBBS reported point data aligned with the center of SPECT/CT voxels which match MC. We report: 1) run time of the GBBS, 2) average difference in non-zero absorbed dose voxels, and 3) percent difference in mean absorbed dose to tumor and normal liver. DVHs and isodose curves were visually compared, and the tetrahedral mesh was compared to SPECT/CT. Results: GBBS calculation required 2 minutes. The average difference over all non-zero absorbed dose voxels (N=318,000) was −0.2 Gy. GBBS mean doses agreed within 1% of MC for both normal liver (39.1 Gy vs 39.3 Gy) and tumor (317 Gy vs 320 Gy). Adaptive meshing allowed sufficient material assignment and yielded smaller tetrahedrons near high activity regions and bone. DVHs had excellent visual agreement. Conclusions: GBBS with adaptive meshing is practical for fast and accurate clinical nuclear medicine dosimetry. Research reported in this publication was supported by the National Cancer Institute of the National Institutes of Health under Award Number R01CA138986. The content is solely the responsibility of the authors and does not necessarily represent the official views of the National Institutes of Health.

The absorbed dose to blood from blood-borne activity

The radiation absorbed dose to blood and organs from activity in the blood is relevant for nuclear medicine dosimetry and for research in biodosimetry. The present study provides coefficients for the average absorbed dose rates to the blood from blood-borne activity for radionuclides frequently used in targeted radiotherapy and in PET diagnostics. The results were deduced from published data for vessel radius-dependent dose rate coefficients and reasonable assumptions on the blood-volume distribution as a function of the vessel radius. Different parts of the circulatory system were analyzed separately. Vessel size information for heart chambers, aorta, vena cava, pulmonary artery, and capillaries was taken from published results of morphometric measurements. The remaining blood not contained in the mentioned vessels was assumed to reside in fractal-like vascular trees, the smallest branches of which are the arterioles or venules. The applied vessel size distribution is consistent with recommendations of the ICRP on the blood-volume distribution in the human. The resulting average absorbed dose rates to the blood per nuclear disintegration per milliliter (ml) of blood are (in 10−11 Gy·s−1·Bq−1·ml) Y-90: 5.58, I-131: 2.49, Lu-177: 1.72, Sm-153: 2.97, Tc-99m: 0.366, C-11: 4.56, F-18: 3.61, Ga-68: 5.94, I-124: 2.55. Photon radiation contributes 1.1–1.2·10−11 Gy·s−1·Bq−1·ml to the total dose rate for positron emitters but significantly less for the other nuclides. Blood self-absorption of the energy emitted by ß-particles in the whole blood ranges from 37% for Y-90 to 80% for Tc-99m. The correspondent values in vascular trees, which are important for the absorbed dose to organs, range from 30% for Y-90 to 82% for Tc-99m.

The Use of Microcontrollers by Scientists in Experimental Research Applications

If the microcontrollers are devices quite easy to use for experts in electronics, a scientist devoted to different research topics from digital electronics is generally totally unable to develop and use these complex components. Most of the time, the researcher is looking for simple devices able to control different parameters during his experimentation. He is obliged in this case to request the intervention of an expert in electronics to develop such a device and the result is sometimes different from the precise initial requirement. Nowadays, it is possible to find on the market, simple units able to satisfy the requirements of the scientist and easy enough to be manipulated by him in specific applications. The purpose of this paper is to present some of these devices and to compare their possibilities and the ways to use them. These devices are compared in optically stimulated dosimetry with applications found in external dosimetry, interventional radiology, nuclear medicine and space dosimetry.

WE‐E‐BRE‐01: An Image‐Based Skeletal Dosimetry Model for the ICRP Reference Adult Female – Internal Electron Sources

Purpose:Limitations seen in previous skeletal dosimetry models, which are still employed in commonly used software today, include the lack of consideration of electron escape and cross‐fire from cortical bone, the modeling of infinite spongiosa, the disregard of the effect of varying cellularity on active marrow self‐irradiation, and the lack of use of the more recent ICRP definition of a 50 micron surrogate tissue region for the osteoprogenitor cells – shallow marrow. These limitations were addressed in the present dosimetry model.Methods:Electron transport was completed to determine specific absorbed fractions to active marrow and shallow marrow of the skeletal regions of the adult female. The bone macrostructure was obtained from the whole‐body hybrid computational phantom of the UF series of reference phantoms, while the bone microstructure was derived from microCT images of skeletal region samples taken from a 45 year‐old female cadaver. The target tissue regions were active marrow and shallow marrow. The source tissues were active marrow, inactive marrow, trabecular bone volume, trabecular bone surfaces, cortical bone volume and cortical bone surfaces. The marrow cellularity was varied from 10 to 100 percent for active marrow self‐irradiation. A total of 33 discrete electron energies, ranging from 1 keV to 10 MeV, were either simulated or modeled analytically.Results:The method of combining macro‐ and microstructure absorbed fractions calculated using MCNPX electron transport was found to yield results similar to those determined with the PIRT model for the UF adult male in the Hough et al. study.Conclusion:The calculated skeletal averaged absorbed fractions for each source‐target combination were found to follow similar trends of more recent dosimetry models (image‐based models) and did not follow current models used in nuclear medicine dosimetry at high energies (due to that models use of an infinite expanse of trabecular spongiosa).

SU-D-19A-02: Electron and Photon Absorbed Fractions for Tumors of Varying Sizes and Compositions

Purpose: To calculate absorbed fractions for mono-energetic photons and electrons in tumors of varying compositions using Monte Carlo simulations in MCNPX. Although tumor dosimetry has been previously investigated, these studies are very limited as they only consider absorbed fractions for soft-tissue tumors. Methods: The tumors were modeled as spheres with radii ranging from 0.10 cm to 6.0 cm and with compositions varying from 100% soft tissue to 100% bone. The energies of both the photons and electrons were varied from 10 keV to 10 MeV and were homogenously distributed throughout the tumor volume. Furthermore, this investigation addresses the issue of spherical versus elliptical tumors. Both prolate and oblate spheroid tumors of different compositions were modeled, and absorbed fractions were calculated for various electron and photon energies. Results: The data clearly shows an absorbed fraction dependence on tumor composition. For example, a soft-tissue model for a 3 MeV electron emitted in a 1 cm diameter bone tumor would have an 83% error, and this same assumption for a 500 keV photon would yield a 74% error. Ultimately, empirical fits were created for each of the five material compositions in order to facilitate the absorbed fraction calculation, requiring only the tumor sizemore » and particle energy. Furthermore, the data shows that absorbed fractions for moderate spheroids can be well approximated by spherical tumors of equal mass to within 8%, but in the extreme cases where the spheroid resembles more of a disk, the errors can be as high as 30%. Conclusion: This comprehensive data set is most valuable for nuclear medicine dosimetry because it incorporates particle type, particle energy, tumor size, and tumor composition. Although mono-energetic particles were modeled, absorbed fractions and S-values may be calculated for any radionuclide via linear interpolation, as long as the particle energies or spectra are known.« less

Self-irradiation of the blood from selected nuclides in nuclear medicine

Nuclear medicine dosimetry and research in biodosimetry often require the knowledge of the absorbed dose to the blood. This study provides coefficients for the absorbed dose rates to the blood related to the activity concentration in the blood as a function of the vessel radius for radionuclides commonly used in targeted radiotherapy and in PET-diagnostics: C-11, F-18, Ga-68, Y-90, Tc-99 m, I-124, I-131, and Lu-177. The energy deposition patterns after nuclear disintegrations in blood vessel lumina (cylinders homogeneously filled with blood) with radii from 0.01 to 25.0 mm were simulated with the Monte-Carlo radiation transport code MCNPX. An additional contribution from photon radiation from activity in blood in the remainder of the body was taken into account based on a reasonable blood distribution model. The fraction of energy absorbed from non-penetrating radiation in the blood is low in thin blood vessels but approaches the total energy emitted by particles with increasing lumen radius. For photon radiation, irradiation to blood in small vessels is almost completely due to radioactive decays in distant blood distributed throughout the body, whereas the contribution from activity in the vessel becomes dominant for lumen radii exceeding 13 mm. The dependences of the absorbed dose rates on the lumen radius can be described with good accuracy by empirical functions which can be used to determine the absorbed doses to the blood and to the surrounding tissue.

The Direct-Zeff software for direct calculation of mass attenuation coefficient, effective atomic number and effective electron number

Determination of the mass attenuation coefficient, μ/ρ, the effective atomic number, Zeff, and the effective electron number, Neff, is very important in the fields of nuclear diagnostics, radiation protection, nuclear medicine and radiation dosimetry. In this work, the Direct-Zeff software was developed for the computation of the mass attenuation coefficient, the effective atomic number and the effective electron number per unit mass in the energy range 1keV–100GeV. The values of the Zeff, Neff and μ/ρ can be determined for total photon interaction with and without coherent interaction as well as partial photon interactions such as coherent scattering, incoherent scattering, photoelectric absorption and pair production by using the Direct-Zeff software. The accuracy of the Direct-Zeff software has been demonstrated by comparing the calculated data and the experimental values for the various materials. The Direct-Zeff software can be freely obtained by contacting with the authors.

Use of the FLUKA Monte Carlo code for 3D patient-specific dosimetry on PET-CT and SPECT-CT images**Based on work presented at the Third European Workshop on Monte Carlo Treatment Planning (Seville, 15–18 May 2012).

Patient-specific absorbed dose calculation for nuclear medicine therapy is a topic of increasing interest. 3D dosimetry at the voxel level is one of the major improvements for the development of more accurate calculation techniques, as compared to the standard dosimetry at the organ level. This study aims to use the FLUKA Monte Carlo code to perform patient-specific 3D dosimetry through direct Monte Carlo simulation on PET-CT and SPECT-CT images. To this aim, dedicated routines were developed in the FLUKA environment. Two sets of simulations were performed on model and phantom images. Firstly, the correct handling of PET and SPECT images was tested under the assumption of homogeneous water medium by comparing FLUKA results with those obtained with the voxel kernel convolution method and with other Monte Carlo-based tools developed to the same purpose (the EGS-based 3D-RD software and the MCNP5-based MCID). Afterwards, the correct integration of the PET/SPECT and CT information was tested, performing direct simulations on PET/CT images for both homogeneous (water) and non-homogeneous (water with air, lung and bone inserts) phantoms. Comparison was performed with the other Monte Carlo tools performing direct simulation as well. The absorbed dose maps were compared at the voxel level. In the case of homogeneous water, by simulating 108 primary particles a 2% average difference with respect to the kernel convolution method was achieved; such difference was lower than the statistical uncertainty affecting the FLUKA results. The agreement with the other tools was within 3–4%, partially ascribable to the differences among the simulation algorithms. Including the CT-based density map, the average difference was always within 4% irrespective of the medium (water, air, bone), except for a maximum 6% value when comparing FLUKA and 3D-RD in air. The results confirmed that the routines were properly developed, opening the way for the use of FLUKA for patient-specific, image-based dosimetry in nuclear medicine.

WE‐C‐144‐01: The Management of Imaging Procedure Dose 2: Nuclear Medicine

Wesley Bolch: Nuclear Medicine Dose IndicesA main clinical application of nuclear medicine is that of functional imaging of normal and diseased tissue, and the localization of malignant tissue and its potential metastatic spread. In these applications, the amount of administered activity is such that the absorbed dose to both imaged and non‐imaged tissues are typically very low and thus stochastic risks of cancer induction are greatly outweighed by the diagnostic benefit of the imaging procedure. Nevertheless, these tissues doses and their stochastic risks should be quantified for each patient, and placed in context of both their cumulative values received over multiple imaging sessions, and of doses and risks received by other diagnostic imaging procedures they may have (fluoroscopy and computed tomography, for example). The role of internal dosimetry in diagnostic nuclear medicine is thus to provide the basis for stochastic risk quantification. Once this risk is quantified, it may be used to optimize the amount of administered activity in order to maximize image quality while minimizing patient risk. This optimization process is of particular importance for pediatric patients owing to their enhanced organ radiosensitivities and years over which any stochastic effects may become manifest. This optimization should consider, as much as possible, patient age, gender, and body morphometry, and pharmacokinetics, along with all available image acquisition and processing techniques. Unlike other forms of diagnostic imaging, for which dose indices are readily measured, only the administered radioactivity is typically available for “dose tracking”. In this course, we will review data sources for organ and effective dose per unit administered activity for the more common molecular imaging radiopharmaceuticals. Particular attention will be given to sources of individual variability in both organ and effective dose attributed to both physiological and anatomical variations among patients.Learning Objectives:1. Identify the more common radiopharmaceuticals used in functional imaging of normal and diseased tissues.2. Demonstrate understanding of the parameters needed to estimate tissue dose during nuclear medicine imaging and therapy.3. Identify fundamental data sources for organ and effective dose per unit administered activity.4. Demonstrate understanding of the physiological and anatomic sources of individual variability in organ and effective dose per unit administered activity.5. Identify key features of new generation anatomical models that can reduce dose uncertainties through improved matching of patient body morphometry.Fred Fahey: Tracking Doses in the Pediatric PopulationA review of methods for estimating patient dose from pediatric nuclear medicine procedures is presented. Included is a discussion of hybrid imaging and factors that affect radiation dose from CT in PET/CT procedures. Consideration for choice of radionuclide and administered activity for various procedures are detailed, with a focus on the pediatric population. Technological advances allowing the further reduction in radiation dose is also discussed.Learning Objectives:1. List three considerations in estimating the radiation dose from pediatric nuclear medicine.2. Discuss three factors that affect the radiation dose from the CT component of hybrid imaging.3. Describe three factors that can affect the appropriate choice of administered activity for a nuclear medicine study.4. List 2 advances that may lead to further reduction in the administered activity in pediatric nuclear medicine.

IAEA Support to Medical Physics in Nuclear Medicine

Through its programmatic efforts and its publications, the International Atomic Energy Agency (IAEA) has helped define the role and responsibilities of the nuclear medicine physicist in the practice of nuclear medicine. This paper describes the initiatives that the IAEA has undertaken to support medical physics in nuclear medicine. In 1984, the IAEA provided guidance on how to ensure that the equipment used for detecting, imaging, and quantifying radioactivity is functioning properly (Technical Document [TECDOC]-137, "Quality Control of Nuclear Medicine Instruments"). An updated version of IAEA-TECDOC-137 was issued in 1991 as IAEA-TECDOC-602, and this included new chapters on scanner-computer systems and single-photon emission computed tomography systems. Nuclear medicine physics was introduced as a part of a project on radiation imaging and radioactivity measurements in the 2002-2003 IAEA biennium program in Dosimetry and Medical Radiation Physics. Ten years later, IAEA activities in this field have expanded to cover quality assurance (QA) and quality control (QC) of nuclear medicine equipment, education and clinical training, professional recognition of the role of medical physicists in nuclear medicine physics, and finally, the coordination of research and development activities in internal dosimetry. As a result of these activities, the IAEA has received numerous requests to support the development and implementation of QA or QC programs for radioactivity measurements in nuclear medicine in many Member States. During the last 5 years, support was provided to 20 Member States through the IAEA's technical cooperation programme. The IAEA has also supported education and clinical training of medical physicists. This type of support has been essential for the development and expansion of the Medical Physics profession, especially in low- and middle-income countries. The need for basic as well as specialized clinical training in medical physics was identified as a priority for healthcare providers in many countries. The IAEA's response to meet the increasing needs for training has been 2-folds. Through its regular program, a priority is given to the development of standardized syllabi and education and clinical training guides. Through its technical cooperation programme, support is given for setting up national medical physics education and clinical training programs in countries. In addition, fellowships are granted for professionals working in the field for specialized training, and workshops are organized at the national and regional level in specialized topics of nuclear medicine physics. So as to support on-the-job training, the IAEA has also setup a gamma camera laboratory in Seibersdorf, Austria. The laboratory is also equipped with QC tools and equipments, and radioisotopes are procured when training events are held. About 2-3 specialized courses are held every year for medical physicists at the IAEA gamma camera laboratory. In the area of research and development, the IAEA supports, through its coordinated research projects, new initiatives in quantitative nuclear medicine and internal dosimetry. The future of nuclear medicine is driven by advances in instrumentation, by the ever increasing availability of computing power and data storage, and by the development of new radiopharmaceuticals for molecular imaging and therapy. Future developments in nuclear medicine are partially driven by, and will influence, nuclear medicine physics and medical physics. To summarize, the IAEA has established a number of programs to support nuclear medicine physics and will continue to do so through its coordinated research activities, education and training in clinical medical physics, and through programs and meetings to promote standardization and harmonization of QA or QC procedures for imaging and treatment of patients.

Dosimetric evaluation of nanotargeted 188Re-liposome with the MIRDOSE3 and OLINDA/EXM programs

The OLINDA/EXM computer code was created as a replacement for the widely used MIRDOSE3 code for radiation dosimetry in nuclear medicine. A dosimetric analysis with these codes was performed to evaluate nanoliposomes as carriers of radionuclides ((188)Re-liposomes) in colon carcinoma-bearing mice. Pharmacokinetic data for (188)Re-N, N-bis (2-mercaptoethyl)-N',N'-diethylethylenediamine ((188)Re-BMEDA) and (188)Re-liposome were obtained for estimation of absorbed doses in normal organs. Radiation dose estimates for normal tissues were calculated using the MIRDOSE3 and OLINDA/EXM programs for a colon carcinoma solid tumor mouse model. Mean absorbed doses derived from(188)Re-BMEDA and (188)Re-liposome in normal tissues were generally similar as calculated by MIRDOSE3 and OLINDA/EXM programs. One notable exception to this was red marrow, wherein MIRDOSE3 resulted in higher absorbed doses than OLINDA/EXM (1.53- and 1.60-fold for (188)Re-BMEDA and (188)Re-liposome, respectively). MIRDOSE3 and OLINDA have very similar residence times and organ doses. Bone marrow doses were estimated by designating cortical bone rather than bone marrow as a source organ. The bone marrow doses calculated by MIRDOSE3 are higher than those by OLINDA. If the bone marrow is designated as a source organ, the doses estimated by MIRDOSE3 and OLINDA programs will be very similar.

Geant4-Simulations for cellular dosimetry in nuclear medicine

The application of unsealed radionuclides in radiobiological experiments can lead to intracellular radionuclide uptake and an increased absorbed dose. Accurate dose quantification is essential to assess observed radiobiological effects. Due to small cellular dimensions direct dose measurement is impossible. We will demonstrate the application of Monte Carlo simulations for dose calculation. Dose calculations were performed using the Geant4 Monte Carlo toolkit, wherefore typical experimental situations were designed. Dose distributions inside wells were simulated for different radionuclides. S values were simulated for spherical cells and cell monolayers of different diameter. Concomitantly experiments were performed using the PC Cl3 cell line with mediated radionuclide uptake. For various activity distributions cellular survival was measured. We yielded S values for dose distribution inside the wells. Calculated S values for a single cell are in good agreement to S values provided in the literature (ratio 0.87 to 1.07). Cross-dose is up to ten times higher for Y-90. Concomitantly performed cellular experiments confirm the dose calculation. Furthermore the necessity of correct dose calculation was shown for assessment of radiobiological effects after application of unsealed radionuclides. Thereby the feasibility of using Geant4 was demonstrated.

Biokinetics and dosimetry of commonly used radiopharmaceuticals in diagnostic nuclear medicine - a review.

PurposeThe impact on patients’ health of radiopharmaceuticals in nuclear medicine diagnostics has not until now been evaluated systematically in a European context. Therefore, as part of the EU-funded Project PEDDOSE.NET (www.peddose.net), we review and summarize the current knowledge on biokinetics and dosimetry of commonly used diagnostic radiopharmaceuticals.MethodsA detailed literature search on published biokinetic and dosimetric data was performed mostly via PubMed (www.ncbi.nlm.nih.gov/pubmed). In principle the criteria for inclusion of data followed the EANM Dosimetry Committee guidance document on good clinical reporting.ResultsData on dosimetry and biokinetics can be difficult to find, are scattered in various journals and, especially in paediatric nuclear medicine, are very scarce. The data collection and calculation methods vary with respect to the time-points, bladder voiding, dose assessment after the last data point and the way the effective dose was calculated. In many studies the number of subjects included for obtaining biokinetic and dosimetry data was fewer than ten, and some of the biokinetic data were acquired more than 20 years ago.ConclusionIt would be of interest to generate new data on biokinetics and dosimetry in diagnostic nuclear medicine using state-of-the-art equipment and more uniform dosimetry protocols. For easier public access to dosimetry data for diagnostic radiopharmaceuticals, a database containing these data should be created and maintained.Electronic supplementary materialThe online version of this article (doi:10.1007/s00259-011-1904-z) contains supplementary material, which is available to authorized users.

Patient-specific dosimetry in radionuclide therapy

This study presents an attempt to compare individualised palliative treatment absorbed doses, by planar images data and Monte Carlo simulation, in two in vivo treatment cases, one of bone metastases and the other of liver lesions. Medical Internal Radiation Dose schema was employed to estimate the absorbed doses. Radiopharmaceutical volume distributions and absorbed doses in the lesions as well as in critical organs were also calculated by Monte Carlo simulation. Individualised planar data calculations remain the method of choice in internal dosimetry in nuclear medicine, but with the disadvantage of attenuation and scatter corrections lack and organ overlay. The overall error is about 7 % for planar data calculations compared with that using Monte Carlo simulation. Patient-specific three-dimensional dosimetric calculations using single-photon emission computed tomography with a parallel computed tomography study is proposed as an accurate internal dosimetry with the additional use of dose-volume histograms, which express dose distributions in cases with obvious inhomogeneity.

Calculation of electron and isotopes dose point kernels with fluka Monte Carlo code for dosimetry in nuclear medicine therapy

The calculation of patient-specific dose distribution can be achieved by Monte Carlo simulations or by analytical methods. In this study, FLUKA Monte Carlo code has been considered for use in nuclear medicine dosimetry. Up to now, FLUKA has mainly been dedicated to other fields, namely high energy physics, radiation protection, and hadrontherapy. When first employing a Monte Carlo code for nuclear medicine dosimetry, its results concerning electron transport at energies typical of nuclear medicine applications need to be verified. This is commonly achieved by means of calculation of a representative parameter and comparison with reference data. Dose point kernel (DPK), quantifying the energy deposition all around a point isotropic source, is often the one. FLUKA DPKS have been calculated in both water and compact bone for monoenergetic electrons (10-3 MeV) and for beta emitting isotopes commonly used for therapy (89Sr, 90Y, 131I 153Sm, 177Lu, 186Re, and 188Re). Point isotropic sources have been simulated at the center of a water (bone) sphere, and deposed energy has been tallied in concentric shells. FLUKA outcomes have been compared to PENELOPE v.2008 results, calculated in this study as well. Moreover, in case of monoenergetic electrons in water, comparison with the data from the literature (ETRAN, GEANT4, MCNPX) has been done. Maximum percentage differences within 0.8.RCSDA and 0.9.RCSDA for monoenergetic electrons (RCSDA being the continuous slowing down approximation range) and within 0.8.X90 and 0.9.X90 for isotopes (X90 being the radius of the sphere in which 90% of the emitted energy is absorbed) have been computed, together with the average percentage difference within 0.9.RCSDA and 0.9.X90 for electrons and isotopes, respectively. Concerning monoenergetic electrons, within 0.8.RCSDA (where 90%-97% of the particle energy is deposed), FLUKA and PENELOPE agree mostly within 7%, except for 10 and 20 keV electrons (12% in water, 8.3% in bone). The discrepancies between FLUKA and the other codes are of the same order of magnitude than those observed when comparing the other codes among them, which can be referred to the different simulation algorithms. When considering the beta spectra, discrepancies notably reduce: within 0.9.X90, FLUKA and PENELOPE differ for less than 1% in water and less than 2% in bone with any of the isotopes here considered. Complete data of FLUKA DPKS are given as Supplementary Material as a tool to perform dosimetry by analytical point kernel convolution. FLUKA provides reliable results when transporting electrons in the low energy range, proving to be an adequate tool for nuclear medicine dosimetry.

RADAR Commentary: Evolution and Current Status of Dosimetry in Nuclear Medicine

There is a long history of absorbed dose calculations being performed to evaluate the risks involved in the application of radiopharmaceuticals to medical studies. In the 1940s, Marinelli, Quimby, and Hine ([1][1]–[3][2]) published a set of articles that presented a summary and approach to

MO‐F‐110‐09: Development of Computational Lymph Node Models for Pediatric Hybrid Phantoms for Nuclear Medicine Dosimetry

Purpose: To develop computational lymph node models implemented within a series of pediatric hybrid phantoms which can be utilized for nuclear medicine dosimetry Methods: The algorithm for generating lymphatic nodes within adult hybrid phantom which we previously reported was improved to use for International Commission on Radiological Protection (ICRP)‐compliant pediatric hybrid phantoms which include newborn, 1‐ year, 5‐year, 10‐year, and 15‐year male and female phantoms. First, a total of 16 lymphatic node anatomical sites were defined within the pediatric hybrid phantoms: extrathoracic, cervical, thoracic (upper and lower), breast (left and right), mesentery (left and right), axillary (left and right), cubital (left and right), inguinal (left and right), and popliteal (left and right) nodal regions. Second, individual lymphatic node positions are randomly sampled in spherical tissue volume centered at each of the previously selected lymphatic node regions. Third, tissue spheres representing individual lymphatic nodes are then created to populate each nodal region of the phantom, with the normal standard nodal diameter until a total lymphatic node mass is achieved to match age‐dependent total reference masses derived from ICRP Publication 66 and 89 reference data. The lymphatic nodes were generated within a total of 10 pediatric hybrid phantoms and coupled with a Monte Carlo code, MCNPX2.6 to estimate radionuclide S values for I‐131 distributed within different lymphatic node sites. Results: Lymphatic node models of which masses were matched to ICRP reference values were implemented into the pediatric hybrid phantom series. An illustrative set of radionuclide S values for I‐131 were calculated where regional lymphatic nodes are considered to be independent source and target regions of patient. Conclusion: The new lymphatic node model suggested in this study can be utilized to obtain anatomically realistic dosimetry in lymphoma patients treated by radionuclide therapy.

MAGIC-f Gel in Nuclear Medicine Dosimetry: study in an external beam of Iodine-131

MAGIC-f gel applicability in Nuclear Medicine dosimetry was investigated by exposure to a 131I source. Calibration was made to provide known absorbed doses in different positions around the source. The absorbed dose in gel was compared with a Monte Carlo Simulation using PENELOPE code and a thermoluminescent dosimetry (TLD). Using MRI analysis for the gel a R2-dose sensitivity of 0.23 s−1Gy−1was obtained. The agreement between dose-distance curves obtained with Monte Carlo simulation and TLD was better than 97% and for MAGIC-f and TLD was better than 98%. The results show the potential of polymer gel for application in nuclear medicine where three dimensional dose distribution is demanded.

SU-GG-I-64: MicroCT-Based Methods of Assessing Imaging Dose to Active Marrow and Endosteum in CT, Fluoroscopy, and Nuclear Medicine

Purpose: To present newly developed microCT-based methods for assessing radiation dose to radiosensitive tissues of the skeleton within computational patient phantoms for CT, fluoroscopy, and nuclear medicine imaging. Methods and Materials: In this study, 13 skeletal sites were harvested from a 40-year male and a 45-year female cadaver from which 32 samples of trabecular spongiosa were cored from each and imaged under microCT. These images were thresholded and coupled to voxelized models of skeletal sites taken from the UF adult hybrid phantoms. EGSnrc-based Paired Image Radiation Transport was then performed for a series of monogenetic electrons localized in various spongiosa and cortical source tissues. The resultant electron absorbed fractions were used to assess (1) radionuclide S values for nuclear medicine dosimetry, and (2) bone-specific fluence-to-dose response functions for CT and fluoroscopic x-ray dosimetry. Results: Radionuclide S values presented in this study uniquely account for the trabecular microstructure of adult bone, account for the finite size and shape of trabecular spongiosa, and allow for electron cross-fire from particle emissions within the cortical regions of mineral bone. For assessment of bone marrow and endosteum dose during CT imaging and fluoroscopic procedures, a detailed library of photon fluence-to-dose response functions is presented for assessing both regional and skeletal average dose to active marrow and endosteum. Conclusions: Calculations of secondary electron transport in the microstructure of trabecular spongiosa of the adult skeleton reveal an enhancement of dose due to secondary electron disequilibrium for x-rays entering the skeleton at energies below 200 keV. Percent excess dose to active marrow over that predicted under the kerma approximation range from 1–5% in bones such as the pelvis to a high of 25% in the cranium. Percent excess dose to the skeletal endosteum can approach 100% in some bone sites over that predicted by the kerma approximation.

Physical dosimetry and mathematical dose calculation in nuclear medicine: A comparative study

Objectives:This paper addresses a comparison between physical dosimetry and mathematical dose calculation in nuclear medicine.Materials and Methods:Dose rate was calculated by mathematical external dose calculation formula and by physical dosimetry from the surface of 38 adult patients’ body referred to nuclear medicine department. Results of the methods were compared and correlation and regression tests were also performed.Results:Although the physical dosimetry data in this study are in good consistency with other researches, they are much lower than the results of mathematical dose calculation formula. The correlation coefficient between measured dose rate with calculated values derived by mathematical formula was found to be 0.852 (P value=0.148).Conclusion:It seems that physical dosimetry data are more accurate than the results of mathematical dose calculation. In case of using mathematical dose calculation formula, other correction factors should be considered and applied for getting reliable data.