ORNL Phantom Research Articles

Using Lithium-6 Filter for Study of Dose Distribution with Max. and Min. Displacement of Prostate Inside the Body by BNCT Method

Physical dose assessment of prostate tumor and adjacent healthy tissues in ORNL, adult male (AM) and KTMAN-2 phantoms was performed using the Monte Carlo transport code. Two different spectra of MIT reactor (without and with lithium filter) were considered. To evaluate the effectiveness of this method in destroying cancer cells for different depths of the prostate, the prostates were moved three times with a distance of 3 cm towards the surface of the body. In addition, the effects of using a lithium filter with different thicknesses on the physical doses of the prostate and adjacent healthy tissues were investigated. It has been observed that decreasing the depth of the prostate increases the physical dose received by the tumor. When the prostate is placed at the maximum depth inside the body, the tissues located in the path of the beam receive the maximum amount of dose. However, the use of a lithium filter reduces damage to healthy tissues, so that when the prostate is near the surface of the body, the lithium filter increases the amount of dose delivered to tumors. In these conditions, the average energy of the penetration distance of epithermal neutrons and neutrons in tissues increases. By changing the location of the prostate gland inside the body, the physical doses of normal tissues did not change, but it was observed that increasing the thickness of the lithium filter significantly reduced the dose of normal tissues.

A dosimetric comparison between ICRP and ORNL phantoms from exposure to 137Cs contaminated soil

The 137Cs soil contamination has detrimental effects on human health. The penetration and distribution of 137Cs radionuclides at different soil depths were investigated. Two sampling campaigns in 2001 and 2018 were performed to determine the 137Cs activity in soil at 11 locations around the city of Kragujevac in central Serbia. The ICRP and ORNL humanoid phantom models were used to determine equivalent and effective doses from the experimentally measured data, and conversion coefficients were obtained from Monte Carlo computations. The results show a general decrease in 137Cs concentration in samples that were taken in 2018. However, variation of 137Cs concentration was observed at different locations that did not match the expected reduction of 137Cs concentration based on its half-life.Considering the computational results, no significant difference between the estimated equivalent dose rate per unit activity of 137Cs (conversion coefficients) between ICRP and ORNL phantoms was observed. In addition, two open-source computer programs for PC and Android devices were developed for users to determine the equivalent and effective doses for their specific cases; this gives the users an easy-to-use interface and portable tools to determine dosimetric quantities. The obtained results and the developed tools would be useful for future investigations related to soil contamination by 137Cs radionuclides that require extensive future studies.

DynamicMC: An Open-source GUI Program Coupled with MCNP for Modeling Relative Dynamic Movement of Radioactive Source and ORNL Phantom in a 3-dimensional Radiation Field.

The present work introduces an open-source graphical user interface (GUI) computer program called DynamicMC. The present program has the ability to generate ORNL phantom input script for the Monte Carlo N-Particle (MCNP) package. The relative dynamic movement of the radiation source with respect to the ORNL phantom can be modeled, which essentially resembles the dynamic movement of source-to-target (i.e., human phantom) distance in a 3-dimensional radiation field. The present program makes the organ-based dosimetry of the human body much easier, as users are not required to write lengthy scripts or deal with any programming that many may find tedious, time consuming, and error prone. In this paper, we have demonstrated that the present program can successfully model simple and complex relative dynamic movements (i.e., those involving rotation of source and human phantom in a 3-dimensional field). The present program would be useful for organ-based dosimetry and could also be used as a tool for teaching nuclear radiation physics and its interaction with the human body.

The second version of InRaDoS, a free computer program for internal dosimetry

The calculation of absorbed and effective doses due to the administration of radiopharmaceuticals needs physical data such as radionuclide's half-life and nuclear decay data. Also, it needs the data on the biodistribution of the radiopharmaceutical in the body. This process is time-consuming, and there is a possibility of human error in the calculations. Therefore, computer programs were developed to facilitate and speed up the dose calculation. The purpose of this paper is to report the steps taken for the development of the second version of InRaDoS. InRaDoS 2 is a free computer program for internal dosimetry. This program was developed using the tkinter module of Python programming language. The dose calculations were based on the MIRD method and only for adults. The user can calculate the doses for revised ORNL or ICRP adult reference phantoms. For now, 98 and 29 radiopharmaceuticals (138 and 42 biokinetic models) are available for the revised ORNL and ICRP adult reference phantoms, respectively. The results of the InRaDoS 2 show good agreement with the published data and free available software. InRaDoS 2 is freely available for download on GitHub(https://github.com/farzinsafarnejad/InRaDoS-2).

Photon-specific absorbed fraction estimates in stylized ORNL and voxelized ICRP adult male phantoms using a new developed Geant4-based code "DoseCalcs": a validation study.

When a radiotracer is injected into a patient's body as part of a nuclear medicine investigation, the entire body is exposed to the ionizing radiation emitted, which can cause biological damage. Therefore, it is important to predict the internal radiation dose to properly balance the advantages of radiological examinations. Currently, various Monte Carlo tools, such as MCNP, Geant4, and GATE, are available to estimate internal radiation dosimetry-related quantities, such as S values (S) and specific absorbed fractions (SAF). Such codes make physics easier for physicists who are experienced with computer programming; however, programming and/or simulation inputs remain a time-consuming and intensive task. In this study, we present a newly developed Geant4-based code for internal dosimetry calculations, namely "DoseCalcs". To assess the performance of the geometrical methods and computational capabilities of our developed tool, we used the GDML, TEXT, STL, and C++ methods to model the ORNL adult phantom, and a voxel-based structure to construct the ICRP adult male. SAFs in the ORNL and ICRP adult male phantoms for eight discrete mono-energetic photons with energies ranging from 0.01 to 2 MeV are calculated with DoseCalcs and compared to ORNL and OpenDose reference data. The two phantoms showed good agreement with both references, which indicates the accuracy of DoseCalcs for subsequent use in estimating internal dosimetry quantities using a variety of geometrical methods.

Organ dose in cardiac dual-energy computed tomography: a Monte Carlo study.

Dual-energy computed tomography (DECT) has appeared as a novel approach with the aim of evaluating artery-related diseases. With the advent of DECT, concerns have been raised about the induction of diseases such as cancer due to high radiation exposure of patients. Therefore, the dose received by patients in DECT should be considered. The parameter most commonly used for patient dosimetry is the effective dose (ED). The purpose of this study is to model and validate a DECT scanner by a developed MCNP Monte Carlo code and to calculate the organ doses, the ED, and the conversion factor (k-factor) used in determining ED in the cardiac imaging protocol. To validate the DECT scanner simulation, a standard dosimetry body phantom was modeled in two radiation modes of single energy CT and DECT. The results of simulated CT dose index (CTDI) were compared with those of ImPACT or measurement data. Then dosimetry phantom was replaced by the male and female ORNL phantoms and the organ doses were calculated. The organ doses were also calculated by ImPACT dose software. In the initial validation stage, the minimum and maximum observed relative differences between results of MNCP simulation and measured were 2.77% and 5.79% for the central CTDI and 1.91% and 5.83% for the averaged peripheral CTDI, respectively. The mean ED of simulation and the ImPACT were 3.23 and 5.55mSv/100mAs, and the mean k-factor was 0.016 and 0.032mSvmGy-1cm-1 in the male and female phantoms, respectively. The k-factor obtained for males is close to the currently used k-factor, but the k-factor for females is almost twice.

Evaluation of Dose Distribution in Lung Tumor Radiotherapy with Boron Neutron Capture Therapy

Introduction: Boron neutron capture therapy (BNCT) is recommended to treat glioblastoma. It is well known that neutrons are more effective treatment than photons to treat hypoxic tumors due to interaction with the nucleus and production of heavy particles. Objective: To evaluate the suitability of BNCT for treating lung cancer, neutron dose distributions was calculated in lung tumor volume and in peripheral organs at risk (OARs).Materials and Methods: Dose distribution to treat lung cancer was calculated by MCNPX code. An elliptical tumor with volume of 27cm3 was centered in the left lung of ORNL phantom and was irradiated with a rectangular field of neutron. Recommended neutron spectrums of MIT and CNEA-MEC were used as a neutron source. The tumor was loaded with different concentrations of Boron 0, 10, 30 and 60 ppm to evaluate the delivered dose to OARs. Results: neutron absorbed dose rate in the tumor was 2.2×10-3, 2.6×10-3, 3.4×10-3 and 4.7×10-3 Gy/s for boron concentrations of 0, 10, 30 and 60 ppm, respectively in MIT. Moreover, the similar results in CNEA-MEC was 1.2×10-3, 1.6×10-3, 2.5×10-3 and 3.7×10-3 Gy/s. Among all, heart absorbed the maximum neutron dose rate of 1.7×10-4 and 1.6×10-4 Gy/s in MIT and CNEA, respectively. For all energy’s bins of spectrums, the neutrons flux is decreased as it penetrates the lung. Conclusion: The implemented model was successful in calculating the dose to organs using BNCT. An Increase in boron concentrations in tumor results in an increase in the absorbed doses while dose uniformity deteriorates. Results showed that the MIT source is well suited to treat deep lung tumors while maintaining the OARs’ dose within the threshold dose.

Estimation of the Effective Dose of Radiation Workers: Optimization Based on the Weight Percentile

Radiation workers might be exposed to polyenergetic photon radiation beams at different directions in their working environments. In this regard, their effective dose (E) should be accurately estimated using a two-dosimeter algorithm (TDA), based on the measurements of two thermoluminescent dosimeters (TLDs) or film badges that are mounted on the front and back of the body. However, considering different human anatomies, radiation workers may have a variety of weight percentiles. This work sought to find whether TDA obtained for the reference weight percentile (50) can be used for higher weight percentiles (including; 65, 75, 85, and 95). MCNPX was used to simulate different weight percentiles on the revised ORNL phantom by adding extra layers of muscle and adipose on the torso. Then front and back TLD responses were calculated for external beam photon energies of 40 keV to 10 MeV in different irradiation geometries. The results revealed that E value declines with increasing the weight percentile. In this study, three TDA were investigated consisting of Eest = 0.73 Rf + 0.53 Rb (73/53), Eest = 0.55 Rf + 0.50 Rb (55/50), and Eest = 0.70 Rf + 0.30 Rb (70/30). The ratio of Eest/E was calculated for each TDA in different energy bins and weight percentiles. Results obtained using the 55/50 and 70/30 showed higher underestimation for most of the energy bins, especially for PA and AP geometries. Compared to these two TDA, the 73/53 algorithm resulted in higher overestimation for RLAT and LLAT geometries for the same energy bins. Variation of the algorithms showed a similar trend for the studied weight percentiles. To conclude, results obtained by TDA for the 50% weight percentile are applicable for weight percentiles >50%.

Measuring and Monte Carlo Modelling of X-Ray and Gamma-Ray Attenuation in Personal Radiation Shielding Protective Clothing.

A collection of personal protective equipment (PPE), suitable for use in case of accident in nuclear facilities or radiological emergencies, was gathered at the National Institute for Nuclear, Chemical and Biological Protection, Czech Republic. The shielding characteristics of the various PPE materials were measured via narrow geometry spectral attenuation measurements with point radionuclide sources covering a broad range of photon energies. Photon relative penetration and attenuation for relevant energies of the spectra were the principal experimentally determined quantities for tested PPE. Monte Carlo simulations in the MCNPX™ code were carried out to determine photon attenuation for respective energies in the tested PPE, and the results were compared to those determined experimentally. Energy depositions in a unit volume of an ORNL phantom were simulated in a radioactive aerosols atmospheric environment to determine effective doses both for the whole body and in various organs in the human torso during exposure to different dispersed radioactive aerosols while wearing one of the PPE protecting against X- and gamma-ray. This work aimed to determine the effective dose and its decrease for individual PPE protecting against X- and gamma-ray.

Dosimetry of Nano-Radio-Ytterbium Therapy by MIRD and MCNP methods for humans’ organs

Introduction: Nano radio-pharmaceutical therapy (NRPT) is a new method for solid tumor therapy. The treatment uses a radioactive form of radionuclide encapsulated in the poly amido amine dendrimers. The poly (amidoamine) (PAMAM) dendrimers have attracted attentions for cancer treatment by their characteristics of targeted drug carriers, delivery agents, and imaging agents in human systems. We have reported the preparation of dendrimer encapsulated ytterbium-175 radio- nanoparticles and its biodistribution in tumor bearing rats.175Yb (T1/2=4.2 days), decays to stable 175Lu with a β- emitter with 470 keV maximum energy (86.5%) and γ photons of 113 keV (1.9%), 282keV (3.1%) and 396 keV (6.5%) that are appropriate for imaging. This paper aims at comparing dosimetric assessments in human performed with Monte Carlo codes and MIRD based on the experimental results of biodistribution of dendrimer encapsulated ytterbium-175.Materials and Methods: All chemical materials including, Ytterbium (III) oxide (Yb2O3), PAMAMG5-NH2 dendrimer in 5% methanol solution and HNO3, Sodium boro- hydride (NaBH4) were purchased from Sigma Aldrich Chemical Co. USA and Merck, Germany. For biodistribution study 20 female Balb/c mice were purchased from Pasteur Institute of Iran. To estimate the absorbed dose by both methods, MCNP and MIRD, the cumulated activity in source organs were calculated by the percentage of injected dose in humans’ organs. The residence times (τ) in the source organs were obtained by integration of respective fit functions, on biokinetics curve, after accounting for the physical decay of the 175Yb. The dose calculation was done for a certain group of organs of human following the MIRD technique. To simulate the problem with MCNP, the ORNL phantom was used. The desired result for this study is absorbed doses of vital organs (liver, spleen, lung and kidney). Results: The maximum uptake of radio-compound are in the liver, lung and spleen. The biodistribution shows the characteristics of nanoparticles such as size and surface hydrophobicity that determine the amount of adsorbed radiopharmaceuticals in organs. The initial fast distribution of radiotracer is throughout liver, lung, and Bone, with slower accumulation in blood and spleen. The result of 2 methods (MIRD versus MCNPX) reveals that MIRD underestimate the absorbed dose for bladder, bone, lung, and ovaries while overestimate for liver, muscle and spleen. In this study the absorbed dose from 175Yb- PAMAM estimated by MCNPX for liver, lung, spleen, kidney and bone are 1.266, 8.081E-01, 8.347E-01, 3.979E-02 and 1.706E-02 mGy/MBq respectively Conclusion: Owing to the stability of PAMAM encapsulated Yb-175 and the size of nano- particle the concentrations are mostly in liver and lungs. State of- the-art dosimetry depends on the duration of the biokinetics of the radiopharmaceutical and a calculation of residence times including an analysis of the errors associated with the respective calculation that aimed in this study. The results showed that this nano-radiopharmaceutical has potential of application for liver and lung tumors.

Investigation the potential of Boron neutron capture therapy (BNCT) to treat the lung cancer

Introduction: Boron neutron capture therapy (BNCT) is recommended to treat the glioblastoma tumor. It is well known that neuron beams are more effective treatment than photon beams to treat hypoxia tumors due to interaction of neutron with nucleus and production of heavy particles such as 7Li and alpha particle. In this study to evaluate the suitability of BNCT for treating of lung cancer, the dose distributions of neutron beam were calculated in lung tumor volume and in peripheral organs at risk (OARs). Materials and Methods: Dose distribution in Boron neutron capture therapy to treat lung cancer was calculated by MCNPX (2.6.0) code. A 3×3×3 cm3 tumor was located in left lung of ORNL phantom and was irradiated with a rectangular field of neutron positioned at surface source distance (SSD) of 10 cm. recommended spectrum of MIT (Massachusetts Institute of Technology) was used. Tumor was loaded with different concentrations of Boron 0, 10, 30 and 60 ppm. Dose delivered to OARs such as heart, spinal cord, right lung … were calculated. Results: The results show that neutron flux significantly decreased followed to penetrate in lung tissue. Neutron flux decreased in all energy bins of irradiated MIT spectrum; maximum fall- off occurred in the range of epithermal energy. Dose distribution was not depended to SSD. The absorbed dose in tumor was 2.16×10-14, 2.6×10-14, 3.44×10-14 and 4.72×10-14 Gy(per one irradiated neutron from source) for boron concentration of 0, 10, 30 and 60 ppm, respectively. From the OARs, the heart tissue absorbed the maximum dose of 1.66×10-15 Gy (per one irradiated neutron from source). Conclusion: Our simulated model was successful to calculated organ doses in BNCT. As the boron concentration in lung tumor increases, absorbed dose increased while dose uniformity trended downward. Our results show that the MIT neutron source is suitable to treat deep lung tumors while OVRs’ dose maintains within the threshold dose.

P089] Calculations of specific absorbed fractions in organs of human body in nuclear medicine due application of 81MKr

Monte Carlo simulations were performed to assess the dose in the treatment of radiopharmaceuticals 81mKr. This radiopharmaceutical is used in treatments in nuclear medicine as an indication for cardiovascular and pulmonary diseases. The aim of this paper was to evaluate the specific absorbed fraction (SAF) when this radiopharmaceutical is incorporated in the lungs. For this purpose, we developed a voxel phantom (thorax) and was compared to the ORNL phantom. All calculations and simulations are done using the MCNP5/X code. Purpose Nuclear medicine uses radioactive isotopes and compounds for diagnosis and therapy. The use of different isotopes depending on the procedure applied to the patient dose distribution. In order to produce the least damage to the patient in therapy and provide good results for diagnostic purposes, the isotope must have a short half-life. Methods In this paper, two types of phantom were used - ORNL mathematical phantoms and voxel phantom. In order to calculate absorbed doses and SAF, a voxel phantom of the thorax was developed. The construction of voxel phantom depends on the quality of digital images of patients obtained during CT or MRI examination. In this paper, a voxel phantom (Thorax model) was obtained using a DICOM set of 108 CT images of the female patient under approved standard protocols. Results The SAF for the selected organ is calculated by the following formula: SAF = E d E i m Where is: the E d is mean energy is deposited throughout the body, E i the primary energy emitted by the source and m is the mass organ/tissue. Calculated SAF values of both phantom models are normalized in relation to the lungs, which provide a comparison of this two models. Conclusion The obtained results of the evaluation of SAF in different organs/tissues, during the incorporation of 81mKr in the lungs, are shown during the process in the scintigraphic examination. The difference in the results obtained by the use of the voxel model and the ORNL mathematical phantom model is between 4.47% and 19.7%.

Monte-Carlo based assessment of MAGIC, MAGICAUG, PAGATUG and PAGATAUG polymer gel dosimeters for ovaries and uterus organ dosimetry in brachytherapy, nuclear medicine and Tele-therapy

Background and objectivesCalculation of 3D dose distribution during radiotherapy and nuclear medicine helps us for better treatment of sensitive organs such as ovaries and uterus. In this research, we investigate two groups of normoxic dosimeters based on meta-acrylic acid (MAGIC and MAGICAUG) and polyacrylamide (PAGATUG and PAGATAUG) for brachytherapy, nuclear medicine and Tele-therapy in their sensitive and critical role as organ dosimeters. MethodsThese polymer gel dosimeters are compared with soft tissue while irradiated by different energy photons in therapeutic applications. This comparison has been simulated by Monte-Carlo based MCNPX code. ORNL phantom–Female has been used to model the critical organs of kidneys, ovaries and uterus. Right kidney is proposed to be the source of irradiation and another two organs are exposed to this irradiation. ResultsEffective atomic numbers of soft tissue, MAGIC, MAGICAUG, PAGATUG and PAGATAUG are 6.86, 7.07, 6.95, 7.28, and 7.07 respectively. Results show the polymer gel dosimeters are comparable to soft tissue for using in nuclear medicine and Tele-therapy. Differences between gel dosimeters and soft tissue are defined as the dose responses. This difference is less than 4.1%, 22.6% and 71.9% for Tele-therapy, nuclear medicine and brachytherapy respectively. The results approved that gel dosimeters are the best choice for ovaries and uterus in nuclear medicine and Tele-therapy respectively. ConclusionsDue to the slight difference between the effective atomic numbers of these polymer gel dosimeters and soft tissue, these polymer gels are not suitable for brachytherapy since the dependence of photon interaction to atomic number, for low energy brachytherapy, had been so effective. Also this dependence to atomic number, decrease for photoelectric and increase for Compton. Therefore polymer gel dosimeters are not a good alternative to soft tissue replacement in brachytherapy.

Comparison of the effectiveness of polymer gel dosimeters (Magic and Pagatug) for organ dose calculation in brachytherapy, nuclear medicine and teletherapy

Purpose: To investigate and compare two polymer gel dosimeters, Magic and Pagatug, as organ dosimeters for 3D measurement of dose distribution in brachytherapy, nuclear medicine and teletherapy. Methods: Magic and Pagatug polymer gels were compared with soft tissue based on irradiation with low energy photons during therapeutic applications. Comparison was simulated using Monte-Carlobased MCNPX code. ORNL phantom–Female was used to model some vital organs (kidneys, ovaries and uterus). The right kidney was proposed to be the source of irradiation and the two organs were exposed to this irradiation. Results : The effective atomic numbers of soft tissue, Magic and Pagatug were 6.86134, 7.07 and 7.2884, respectively. The results showed that Magic and Pagatug, were comparable to soft tissue with regard to application in nuclear medicine and teletherapy. Differences between gel dosimeters and soft tissue were defined as the dose responses. This difference was < 8.1, < 4 and < 76.8 % for teletherapy, nuclear medicine and brachytherapy, respectively. Conclusion: Due to slight differences between the effective atomic numbers of these polymer gel dosimeters and soft tissue, the polymer gels are not suitable for brachytherapy since the photoelectric interaction is dominant for low energy photons, and the interaction relates to Z4. The results demonstrate that the gel dosimeters are best suited for nuclear medicine. Keywords : Magic, Pagatug, Brachytherapy, Nuclear medicine, Teletherapy, Organ dosimetry, Soft tissue

Abstract ID: 225 Development of adjustable model breast using mammographic information for Monte Carlo simulation in nuclear medicine — Impact on dosimetry

For accurate estimation of the breast dose in nuclear medicine examinations, the development of an adjustable sizes model breast for Monte Carlo simulation is important. The development of the model breast for nuclear medicine studies is difficult because the size of the breast must be known. The purpose of this study is to develop the adjustable sizes model breast using mammographic information for Monte Carlo simulation in nuclear medicine and to investigate its impact on dosimetry using the SimDOSE simulation code [1] . This study constructed two sizes of ORNL phantoms with two radionuclides: 18F and 99mTc. The dimension of a breast phantom, a semi-oblate spheroid model, was established from the collected breast parameters of a cohort of patients in mammography [2] to provide a model breast allowing for changing sizes. Breast phantoms with various breast sizes, shapes, compositions, and the glandular distribution were created. The preliminary results showed the ratios of S-values of the breast ranged from 0.94 to 1.01 and the difference caused by breast shapes was insignificant. The organ dose of left and right breast may be considered separately due to in the dose difference between left and right breast. For different glandular distributions, the differences in S-values of the gland self-absorbed dose is about 5–8%. We conclude that the breast size has significant impacts and the glandular distribution has moderate impacts on the S-values of the breast, while the breast shape and compositions cause minor changes in the S-value. Therefore, this study provided the new method to assess the breast dose in Nuclear Medicine.

MCNPX CALCULATIONS OF SPECIFIC ABSORBED FRACTIONS IN SOME ORGANS OF THE HUMAN BODY DUE TO APPLICATION OF 133Xe, 99mTc and 81mKr RADIONUCLIDES.

Monte Carlo simulations were performed to evaluate treatment doses with wide spread used radionuclides 133Xe, 99mTc and 81mKr. These different radionuclides are used in perfusion or ventilation examinations in nuclear medicine and as indicators for cardiovascular and pulmonary diseases. The objective of this work was to estimate the specific absorbed fractions in surrounding organs and tissues, when these radionuclides are incorporated in the lungs. For this purpose a voxel thorax model has been developed and compared with the ORNL phantom. All calculations and simulations were performed by means of the MCNP5/X code.

Dose Conversion Coefficients Based on Taiwanese Reference Phantoms and Monte Carlo Simulations for Use in External Radiation Protection.

Reference phantoms are widely applied to evaluate the radiation dose for external exposure. However, the frequently used reference phantoms are based on Caucasians. Dose estimation for Asians using a Caucasian phantom can result in significant errors. This study recruited 40 volunteers whose body sizes are close to the average Taiwanese population. Magnetic resonance imaging was performed to obtain the organ volume for construction of the Taiwanese reference man (TRM) and Taiwanese reference woman (TRW). The dose conversion coefficients (DCC) resulting from photo beams in anterior-posterior, posterior-anterior, right-lateral, left-lateral, and isotropic irradiation geometries were estimated. In the anterior-posterior geometry, the mean DCC differences among organs between the TRM and ORNL phantom at 0.1, 1, and 10 MeV were 7.3%, 5.8%, and 5.2%, respectively. For the TRW, the mean differences from the ORNL phantom at the three energies were 10.6%, 7.4%, and 8.3%. The DCCs of the Taiwanese reference phantoms and the ORNL phantom presented similar trends in other geometries. The torso size of the phantom and the mass and geometric location of the organ have a significant influence on the DCC. The Taiwanese reference phantoms can be used to establish dose guidelines and regulations for radiation protection from external exposure.

Development and validation of a fast voxel-based dose evaluation system in nuclear medicine

Abstract PET imaging has been widely used in the detection and staging of malignancies and the evaluation of patient-specific dosimetry for PET scans is important in nuclear medicine. However, patient-specific dosimetry can be estimated only by Monte Carlo methods which are usually time-consuming. The purpose of this study is to develop a fast dose evaluation system namely SimDOSE. SimDOSE is a Monte Carlo code embedded in SimSET with a dose scoring routine to record the deposited energy of the photons and electrons. Fluorine-18 is one of the most commonly used radionuclides that decay predominantly by positron emission. Only a 635 keV (Emax) positron and two annihilation photons should be concerned in F-18 radiation dosimetry, hence simulation is relatively simple. To evaluate the effects of resolution, an F-18 point source placed in a 20 cm diameter sphere filled with water was simulated by SimDOSE and GATE v6.1. Grid sizes of 1 mm, 3 mm, and 5 mm were tested and each was simulated with a total of 10 7 decays. The resultant dose distribution functions were compared. Dose evaluation on ORNL phantom was also performed to validate the accuracy of SimDOSE. The grid size of phantom was set as 3 mm and the number of decays was 10 7 . The S -values of liver computed by SimDOSE were compared with GATE and OLINDA (Organ Level INternal Dose Assessment) for 11 C, 15 O, and 18 F.Finally, the CPU time of simulations was compared between SimDOSE and GATE. The dose profiles show the absorption doses located 3 mm outside the center are similar between SimDOSE and GATE. However, 71% (19%) difference of the center dose between SimDOSE and GATE are observed for 1 mm (3 mm) grid. The differences of the profile lie in the assumption in SimDOSE that all kinetic energies of electrons are locally absorbed. The ratios of S values of (SimDOSE/OLINDA) range from 0.95 to 1.11 with a mean value of 1.02±0.043. To compare simulation time from SimDOSE to GATE for calculation of 1 mm, 3 mm, 5 mm gird point source and S values of ORNL phantom are 1.3%, 1.2%, 1.2% and 1.2%, respectively. In conclusion, SimDOSE is an efficient and accurate toolkit to generate patient-specific dose distribution in clinical PET application.

MCNP simulation of the dose distribution in liver cancer treatment for BNC therapy

AbstractThe Boron Neutron Capture Therapy (BNCT) is based on selective uptake of boron in tumour tissue compared to the surrounding normal tissue. Infusion of compounds with boron is followed by irradiation with neutrons. Neutron capture on 10B, which gives rise to an alpha particle and recoiled 7Li ion, enables the therapeutic dose to be delivered to tumour tissue while healthy tissue can be spared. Here, therapeutic abilities of BNCT were studied for possible treatment of liver cancer using thermal and epithermal neutron beam. For neutron transport MCNP software was used and doses in organs of interest in ORNL phantom were evaluated. Phantom organs were filled with voxels in order to obtain depth-dose distributions in them. The result suggests that BNCT using an epithermal neutron beam could be applied for liver cancer treatment.

Improvements on a patient-specific dose estimation system in nuclear medicine examination

The purpose of this paper is to develop a patient-specific dose estimation system in nuclear medicine examination. A dose deposition routine to store the deposited energy of the photons during their flights was embedded in the widely used SimSET Monte Carlo code and a user-friendly interface for reading PET and CT images was developed. Dose calculated on ORNL phantom was used to validate the accuracy of this system. The ratios of S value for (99m)Tc, (18)F and (131)I computed by this system to those obtained with OLINDA for various organs were ranged from 0.93 to 1.18, which were comparable to that obtained from MCNPX2.6 code (0.88-1.22). Our system developed provides opportunity for tumor dose estimation which cannot be known from the MIRD. The radiation dose can provide useful information in the amount of radioisotopes to be administered in radioimmunotherapy.