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DOE OSTI2014

Incorporating single-side sparing in models for predicting parotid dose sparing in head and neck IMRT

Purpose: Sparing of single-side parotid gland is a common practice in head-and-neck (HN) intensity modulated radiation therapy (IMRT) planning. It is a special case of dose sparing tradeoff between different organs-at-risk. The authors describe an improved mathematical model for predicting achievable dose sparing in parotid glands in HN IMRT planning that incorporates single-side sparing considerations based on patient anatomy and learning from prior plan data. Methods: Among 68 HN cases analyzed retrospectively, 35 cases had physician prescribed single-side parotid sparing preferences. The single-side sparing model was trained with cases which had single-side sparing preferences, while the standard model was trained with the remainder of cases. A receiver operating characteristics (ROC) analysis was performed to determine the best criterion that separates the two case groups using the physician's single-side sparing prescription as ground truth. The final predictive model (combined model) takes into account the single-side sparing by switching between the standard and single-side sparing models according to the single-side sparing criterion. The models were tested with 20 additional cases. The significance of the improvement of prediction accuracy by the combined model over the standard model was evaluated using the Wilcoxon rank-sum test. Results: Using the ROC analysis, the best single-side sparing criterion is (1) the predicted median dose of one parotid is higher than 24 Gy; and (2) that of the other is higher than 7 Gy. This criterion gives a true positive rate of 0.82 and a false positive rate of 0.19, respectively. For the bilateral sparing cases, the combined and the standard models performed equally well, with the median of the prediction errors for parotid median dose being 0.34 Gy by both models (p = 0.81). For the single-side sparing cases, the standard model overestimates the median dose by 7.8 Gy on average, while the predictions by the combined model differ from actual values by only 2.2 Gy (p = 0.005). Similarly, the sum of residues between the modeled and the actual plan DVHs is the same for the bilateral sparing cases by both models (p = 0.67), while the standard model predicts significantly higher DVHs than the combined model for the single-side sparing cases (p = 0.01). Conclusions: The combined model for predicting parotid sparing that takes into account single-side sparing improves the prediction accuracy over the previous model.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2015

Modeling the effectiveness of remediation efforts in contaminated urban areas: An EMRAS II Urban Areas Working Group exercise - 15631

The Urban Areas Working Group was part of the IAEA's EMRAS II (Environmental Modelling for Radiation Safety) Programme during 2009-2011. The Working Group's goal was to test and improve the capabilities of models used in assessment of radioactive contamination in urban settings. The 'contaminant transport and countermeasures' exercise was focused on modeling the effectiveness of potential countermeasures or remediation efforts with respect to reduction of human exposures and doses. The exercise was based on a hypothetical situation involving an initial concentration of Co-60 or Pu-239 in air in a specified urban area. Participants were given detailed input information, including descriptions of test locations in two types of urban areas, (1) a business area with buildings and asphalt, and (2) a park area near an apartment town. Several kinds of initial weather conditions (dry, light rain, and heavy rain) were explored. Modeling endpoints included contamination densities, dose rates (from each contributing surface and total), annual and cumulative doses (internal and external) for specified reference individuals, and effectiveness of countermeasures and remediation efforts in terms of dose rate reduction and dose reduction. External doses were emphasized for Co-60 and internal doses for Pu-239. Approaches and model predictions from five models are compared, and explanations for similarities and differences among model predictions are offered. Understanding the effects of different approaches, assumptions, and parameter values is essential to a comparison of results from different models and to understanding both the usefulness and the limitations of any given model. (authors)

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2014

Applicability of ICRP 116 dose conversion coefficients to Korean pediatric population exposed to external photons - Paper 25

In the event of a radiological accident or attack, the estimation of organ dose is crucial to determine the possible health effect to general public. To that end, International Commission on Radiological Protection (ICRP) published a comprehensive set of organ dose conversion coefficients derived from the adult stylized (or mathematical) computational phantoms. More recently, ICRP updated the conversion coefficients based on their new reference computational phantoms, adult male and female voxel phantoms, exposed to a variety of radiation types and irradiation geometries. The adult voxel phantoms represent the organ mass and body dimensions of the Caucasian population. It is known that children generally receive greater radiation dose to organs compared to adults for external irradiation due to their smaller body size. that may provide less shielding to internal organs than adults. However, only a few studies investigated if the organ dose conversion coefficients developed for adults can be used for children and how much dose discrepancy needs to be expected. In this study, we adopted a Korean 7-year-old whole body voxel phantom to calculate a comprehensive organ dose conversion coefficients when the phantom is exposed to external photon irradiation geometries. We also compared the results with the data provided in the ICRP Publication 116, which is based on the ICRP adult voxel phantoms. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2012

Estimation of organ and effective dose due to Compton backscatter security scans

Purpose: To estimate organ and effective radiation doses due to backscatter security scanners using Monte Carlo simulations and a voxelized phantom set. Methods: Voxelized phantoms of male and female adults and children were used with the GEANT4 toolkit to simulate a backscatter security scan. The backscatter system was modeled based on specifications available in the literature. The simulations modeled a 50 kVp spectrum with 1.0 mm-aluminum-equivalent filtration and a previously measured exposure of approximately 4.6 {mu}R at 30 cm from the source. Photons and secondary interactions were tracked from the source until they reached zero kinetic energy or exited from the simulation's boundaries. The energy deposited in the phantoms' respective organs was tallied and used to calculate total organ dose and total effective dose for frontal, rear, and full scans with subjects located 30 and 75 cm from the source. Results: For a full screen, all phantoms' total effective doses were below the established 0.25 {mu}Sv standard, with an estimated maximum total effective dose of 0.07 {mu}Sv for full screen of a male child. The estimated maximum organ dose due to a full screen was 1.03 {mu}Gy, deposited in the adipose tissue of the male child phantom when located 30 cm from the source. All organ dose estimates had a coefficient of variation of less than 3% for a frontal scan and less than 11% for a rear scan. Conclusions: Backscatter security scanners deposit dose in organs beyond the skin. The effective dose is below recommended standards set by the Health Physics Society (HPS) and the American National Standards Institute (ANSI) assuming the system provides a maximum exposure of approximately 4.6 {mu}R at 30 cm.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2016

SU-G-201-16: Thermal Imaging in Source Visualization and Radioactivity Measurement for High Dose Rate Brachytherapy

Purpose: High Dose Rate (HDR) brachytherapy poses a special challenge to radiation safety and quality assurance (QA) due to its high radioactivity, and it is thus critical to verify the HDR source location and its radioactive strength. This study demonstrates a new method for measuring HDR source location and radioactivity utilizing thermal imaging. A potential application would relate to HDR QA and safety improvement. Methods: Heating effects by an HDR source were studied using Finite Element Analysis (FEA). Thermal cameras were used to visualize an HDR source inside a plastic applicator made of polyvinylidene difluoride (PVDF). Using different source dwell times, correlations between the HDR source strength and heating effects were studied, thus establishing potential daily QA criteria using thermal imaging Results: For an Ir1?2 source with a radioactivity of 10 Ci, the decay-induced heating power inside the source is ∼13.3 mW. After the HDR source was extended into the PVDF applicator and reached thermal equilibrium, thermal imaging visualized the temperature gradient of 10 K/cm along the PVDF applicator surface, which agreed with FEA modeling. For Ir{sup 192} source activities ranging from 4.20–10.20 Ci, thermal imaging could verify source activity with an accuracy of 6.3% with a dwell time of 10 sec, and an accuracy of 2.5 % with 100 sec. Conclusion: Thermal imaging is a feasible tool to visualize HDR source dwell positions and verify source integrity. Patient safety and treatment quality will be improved by integrating thermal measurements into HDR QA procedures.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2022

Pacific Northwest National Laboratory Annual Site Environmental Report for Calendar Year 2021 (Final Report)

The U.S. Department of Energy requires that all its site facilities develop an annual site environmental report to comply with DOE Order 231.1B Chg 1 and DOE Order 458.1 Admin Chg 4. This report provides a synopsis of calendar year 2021 information related to environmental management performance and compliance efforts at the Pacific Northwest National Laboratory. It summarizes site compliance with federal, state, and local environmental laws, regulations, policies, directives, permits, and Orders, and provides environmental management performance benchmarks and their status to the public, regulatory agencies, community officials, Native American tribes, and public interest groups.

54 ENVIRONMENTAL SCIENCES↗

DOE OSTI2019

Maxine Onsite: Estimating Dose from Chronic Atmospheric Radioactive Releases

Maxine Onsite: Developed to estimate dose to individuals who are working onsite but are not site employees. Follows the methodologies discussed in US NRC Regulatory Guides 1.109 and 1.111 to determine dose from a chronic atmospheric release. Determines the dose given relative air concentrations and deposition concentrations as input. We look at four locations onsite that may be accessed by members of the public. To show compliance with the EPA and DOE public dose limit of 10 mrem/yr, determining who is a member of the public or a worker onsite is necessary. SRS has defined a member of the public for NESHAPS and DOE O 458.1. Maxine Onsite runs on Microsoft Windows Operating System through a Microsoft Excel spreadsheet. Input from the user includes parameters that are variable such as air concentrations, and fixed, such as distance to the receptor and elemental iodine fraction. Maxine Onsite was developed from Maxine which is a program meant to calculate dose to the MEI via three pathways, external, inhalation, and ingestion. Maxine Onsite is meant for looking at the exposure through the external and inhalation pathways only, no local food production onsite. The ways that Maxine Onsite is used to calculate the dose to the MEI includes the following: - Determining the respective air concentration values at each location using distances from the release stack; - Breathing rates for the MEI; - Determining the amount of time the MEI is outside; - Determining the plume travel time. The following relative air and ground deposition values for proper execution of the spreadsheet come from MAXDOSE-SR and are needed to accurately determine the dose to the MEI: relative air concentration, relative air concentration decayed by 2.26 days, relative air concentration decayed and depleted for 8 days, and relative deposition. Test cases were designed, developed, and executed to determine the accuracy of Maxine Onsite, which were verified additionally through hand calculations. Several changes were made for testing, including varying air concentrations, distances, breathing rates, and shielding factors. Each test case was then verified through hand calculations to assure the accuracy of the calculated doses. The results from the test cases, as well as the default parameters, can be used for benchmarking Maxine Onsite in the future. Maxine Onsite performs as expected and produces correct and accurate results at each of the different locations onsite (TRL, B-Area Barracks, SRELCC, and Gunsite 51) for varying test cases. The input required by the user is minimal and includes the results of the dose to the MEI on the first page for ease of access. Future plans for Maxine Onsite include adding in the ingestion pathway and adding in atmospheric dispersion estimates to the spreadsheet.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2022

US Fallout Shelter

Being indoors (sheltering) reduces radiation exposures resulting from nuclear fallout or a power-plant accident. However, most US building types, particularly non-residential buildings, lack quantitative estimates. We provide here a high-level modeling analysis of modern US building protection based on a novel radiation protection building attribute taxonomy, a new building protection model, compilation of prior experimental results, and nationally representative building survey data. This approach provides a consistent, quantitative understanding of US building protection and, when combined with the distribution of people among different buildings, can assist in selecting emergency response strategies. We find that indoor radiation protection varies by orders of magnitude. Most people in non-residential buildings are adequately protected, particularly below ground or in the building center. Residential buildings are less protective and those with lighter weight walls lack adequate, above ground protection. This work also highlights key areas where further investigation will improve the current results. These include the improving the understanding of residential basement protection, additional experimental data on non-residential buildings (particularly schools and industrial buildings), and the frequency of brick veneer and stucco exterior residential buildings.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2014

Portal dosimetry for VMAT using integrated images obtained during treatment

Purpose: Portal dosimetry provides an accurate and convenient means of verifying dose delivered to the patient. A simple method for carrying out portal dosimetry for volumetric modulated arc therapy (VMAT) is described, together with phantom measurements demonstrating the validity of the approach. Methods: Portal images were predicted by projecting dose in the isocentric plane through to the portal image plane, with exponential attenuation and convolution with a double-Gaussian scatter function. Appropriate parameters for the projection were selected by fitting the calculation model to portal images measured on an iViewGT portal imager (Elekta AB, Stockholm, Sweden) for a variety of phantom thicknesses and field sizes. This model was then used to predict the portal image resulting from each control point of a VMAT arc. Finally, all these control point images were summed to predict the overall integrated portal image for the whole arc. The calculated and measured integrated portal images were compared for three lung and three esophagus plans delivered to a thorax phantom, and three prostate plans delivered to a homogeneous phantom, using a gamma index for 3% and 3 mm. A 0.6 cm{sup 3} ionization chamber was used to verify the planned isocentric dose. The sensitivity of this method to errors in monitor units, field shaping, gantry angle, and phantom position was also evaluated by means of computer simulations. Results: The calculation model for portal dose prediction was able to accurately compute the portal images due to simple square fields delivered to solid water phantoms. The integrated images of VMAT treatments delivered to phantoms were also correctly predicted by the method. The proportion of the images with a gamma index of less than unity was 93.7% ± 3.0% (1SD) and the difference between isocenter dose calculated by the planning system and measured by the ionization chamber was 0.8% ± 1.0%. The method was highly sensitive to errors in monitor units and field shape, but less sensitive to errors in gantry angle or phantom position. Conclusions: This method of predicting integrated portal images provides a convenient means of verifying dose delivered using VMAT, with minimal image acquisition and data processing requirements.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2014

ITAR: A modified TAR method to determine depth dose distribution for an ophthalmic device that performs kilovoltage x-ray pencil-beam stereotaxy

Purpose: New technology has been developed to treat age-related macular degeneration (AMD) using 100 kVp pencil-beams that enter the patient through the radio-resistant sclera with a depth of interest between 1.6 and 2.6 cm. Measurement of reference and relative dose in a kilovoltage x-ray beam with a 0.42 cm diameter field size and a 15 cm source to axis distance (SAD) is a challenge that is not fully addressed in current guidelines to medical physicists. AAPM's TG-61 gives dosimetry recommendations for low and medium energy x-rays, but not all of them are feasible to follow for this modality. Methods: An investigation was conducted to select appropriate equipment for the application. PTW's Type 34013 Soft X-Ray Chamber (Freiburg, Germany) and CIRS's Plastic Water LR (Norfolk, VA) were found to be the best available options. Attenuation curves were measured with minimal scatter contribution and thus called Low Scatter Tissue Air Ratio (LSTAR). A scatter conversion coefficient (C{sub scat}) was derived through Monte Carlo radiation transport simulation using MCNPX (LANL, Los Alamos, NM) to quantify the difference between a traditional TAR curve and the LSTAR curve. A material conversion coefficient (C{sub mat}) was determined through experimentation to evaluate the difference in attenuation properties between water and Plastic Water LR. Validity of performing direct dosimetry measurements with a source to detector distance other than the treatment distance, and therefore a different field size due to a fixed collimator, was explored. A method—Integrated Tissue Air Ratio (ITAR)—has been developed that isolates each of the three main radiological effects (distance from source, attenuation, and scatter) during measurement, and integrates them to determine the dose rate to the macula during treatment. Results: LSTAR curves were determined to be field size independent within the range explored, indicating that direct dosimetry measurements may be performed with a source to detector distance of 20 cm even though the SAD is 15 cm during treatment. C{sub scat} varied from 1.102 to 1.106 within the range of depths of interest. The experimental variance among repeated measurements of C{sub mat} was larger than depth dependence, so C{sub mat} was estimated as1.019 for all depths of interest. Conclusions: Equipment selection, measurement techniques, and formalism for the determination of dose rate to the macula during stereotaxy for AMD have been determined and are strongly recommended by the authors of this paper to be used by clinical medical physicists.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2015

Electron dose distributions caused by the contact-type metallic eye shield: Studies using Monte Carlo and pencil beam algorithms

A metallic contact eye shield has sometimes been used for eyelid treatment, but dose distribution has never been reported for a patient case. This study aimed to show the shield-incorporated CT-based dose distribution using the Pinnacle system and Monte Carlo (MC) calculation for 3 patient cases. For the artifact-free CT scan, an acrylic shield machined as the same size as that of the tungsten shield was used. For the MC calculation, BEAMnrc and DOSXYZnrc were used for the 6-MeV electron beam of the Varian 21EX, in which information for the tungsten, stainless steel, and aluminum material for the eye shield was used. The same plan was generated on the Pinnacle system and both were compared. The use of the acrylic shield produced clear CT images, enabling delineation of the regions of interest, and yielded CT-based dose calculation for the metallic shield. Both the MC and the Pinnacle systems showed a similar dose distribution downstream of the eye shield, reflecting the blocking effect of the metallic eye shield. The major difference between the MC and the Pinnacle results was the target eyelid dose upstream of the shield such that the Pinnacle system underestimated the dose by 19 to 28% and 11 to 18% for the maximum and the mean doses, respectively. The pattern of dose difference between the MC and the Pinnacle systems was similar to that in the previous phantom study. In conclusion, the metallic eye shield was successfully incorporated into the CT-based planning, and the accurate dose calculation requires MC simulation.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2016

Dose Mapping from a DD Neutron Generator at the University of Sharjah

The University of Sharjah recently acquired a model DD109.4M neutron generator from Adelphi Technology. The generator is moderated and has three neutron ports which are capable of providing fast and thermal neutrons. This neutron generator is intended to be used for various studies like neutron activation analysis, neutron imaging, and profiling and irradiation effects. It is imperative that appropriate measures are taken in order to minimize the radiation dose to the operators, students, and the public. The resulting dose must not exceed the limits set by the UAE Federal Authority for Nuclear Regulations (FANR) and the International Atomic Energy Agency (IAEA). The objective of this study is to generate a neutron and photon dose map of the generator room and its surrounding using simulations, experimentally verify this dose mapping, and set the operating and safety procedures for the neutron generator. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2015

SU-E-T-534: Level of Residual Radioactivity of Activated Parts of a Decommissioned Cyclotron

Purpose: CTI cyclotron RDS-111 was used at the Hong Kong Sanatorium and Hospital (HKSH) to produce radiopharmaceuticals and radioactive tracers for diagnostic scans between 1999 and 2007. During the operation, some machine components became radioactive by activation. For the safety of staff, decommissioning took place in 2009, two years after the cyclotron had stopped operation. This study investigates the residual radioactivity and radionuclides found in different cyclotron components in 2014 in compliance with the local regulations in Hong Kong for transfer of radioactive waste. Methods: A representative sample of each part was counted using a high-purity germanium detector (manufacturer: ORTECT) for at least four hours. GammaVision, a multichannel analyzer software, was used to identify the radionuclides found in the cyclotron components, as well as the associated activities. A standard library and a Mariscotti peak search algorithm were used to identify the present radionuclides. Only radionuclides with half-life greater than 180 days were considered. Results: Among the components, the Havar target foil has the highest specific activity ((4.6±0.6)×10{sup 2} Bq/g), with Co-60 being the most prominent ((3.8±0.5)×10{sup 2} Bq/g). The total activity of the target foil, however, is still low due to its small mass of 0.04 g. Radioisotopes Mn-54 (46±6 Bq/g), Na-22 (6.8±0.8 Bq/g), Co-57 (7.3±0.9 Bq/g), and Fe-59 (6.0±0.9 Bq/g) have also been detected in the target foil. The target window holder and the vacuum window register a specific activity of 88.3±0.6 Bq/g and 48.6±0.1 Bq/g, respectively. Other components, such as the collimator, the target tube, the valve body and the beamline, are also found with trace amounts of radionuclides. Conclusion: Even seven years after the cyclotron had stopped operation, some components still exhibited residual radioactivity from activation exceeding the IAEA clearance levels. Special consideration for radiological protection may need to be taken in the disposal of the cyclotron parts.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2014

Comparison of patient specific dose metrics between chest radiography, tomosynthesis, and CT for adult patients of wide ranging body habitus

Purpose: Given the radiation concerns inherent to the x-ray modalities, accurately estimating the radiation doses that patients receive during different imaging modalities is crucial. This study estimated organ doses, effective doses, and risk indices for the three clinical chest x-ray imaging techniques (chest radiography, tomosynthesis, and CT) using 59 anatomically variable voxelized phantoms and Monte Carlo simulation methods. Methods: A total of 59 computational anthropomorphic male and female extended cardiac-torso (XCAT) adult phantoms were used in this study. Organ doses and effective doses were estimated for a clinical radiography system with the capability of conducting chest radiography and tomosynthesis (Definium 8000, VolumeRAD, GE Healthcare) and a clinical CT system (LightSpeed VCT, GE Healthcare). A Monte Carlo dose simulation program (PENELOPE, version 2006, Universitat de Barcelona, Spain) was used to mimic these two clinical systems. The Duke University (Durham, NC) technique charts were used to determine the clinical techniques for the radiographic modalities. An exponential relationship between CTDI{sub vol} and patient diameter was used to determine the absolute dose values for CT. The simulations of the two clinical systems compute organ and tissue doses, which were then used to calculate effective dose and risk index. The calculation of the two dose metrics used the tissue weighting factors from ICRP Publication 103 and BEIR VII report. Results: The average effective dose of the chest posteroanterior examination was found to be 0.04 mSv, which was 1.3% that of the chest CT examination. The average effective dose of the chest tomosynthesis examination was found to be about ten times that of the chest posteroanterior examination and about 12% that of the chest CT examination. With increasing patient average chest diameter, both the effective dose and risk index for CT increased considerably in an exponential fashion, while these two dose metrics only increased slightly for radiographic modalities and for chest tomosynthesis. Effective and organ doses normalized to mAs all illustrated an exponential decrease with increasing patient size. As a surface organ, breast doses had less correlation with body size than that of lungs or liver. Conclusions: Patient body size has a much greater impact on radiation dose of chest CT examinations than chest radiography and tomosynthesis. The size of a patient should be considered when choosing the best thoracic imaging modality.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2014

Limitations of the TG-43 formalism for skin high-dose-rate brachytherapy dose calculations

Purpose: In skin high-dose-rate (HDR) brachytherapy, sources are located outside, in contact with, or implanted at some depth below the skin surface. Most treatment planning systems use the TG-43 formalism, which is based on single-source dose superposition within an infinite water medium without accounting for the true geometry in which conditions for scattered radiation are altered by the presence of air. The purpose of this study is to evaluate the dosimetric limitations of the TG-43 formalism in HDR skin brachytherapy and the potential clinical impact. Methods: Dose rate distributions of typical configurations used in skin brachytherapy were obtained: a 5 cm × 5 cm superficial mould; a source inside a catheter located at the skin surface with and without backscatter bolus; and a typical interstitial implant consisting of an HDR source in a catheter located at a depth of 0.5 cm. Commercially available HDR{sup 60}Co and {sup 192}Ir sources and a hypothetical {sup 169}Yb source were considered. The Geant4 Monte Carlo radiation transport code was used to estimate dose rate distributions for the configurations considered. These results were then compared to those obtained with the TG-43 dose calculation formalism. In particular, the influence of adding bolus material over the implant was studied. Results: For a 5 cm × 5 cm{sup 192}Ir superficial mould and 0.5 cm prescription depth, dose differences in comparison to the TG-43 method were about −3%. When the source was positioned at the skin surface, dose differences were smaller than −1% for {sup 60}Co and {sup 192}Ir, yet −3% for {sup 169}Yb. For the interstitial implant, dose differences at the skin surface were −7% for {sup 60}Co, −0.6% for {sup 192}Ir, and −2.5% for {sup 169}Yb. Conclusions: This study indicates the following: (i) for the superficial mould, no bolus is needed; (ii) when the source is in contact with the skin surface, no bolus is needed for either {sup 60}Co and {sup 192}Ir. For lower energy radionuclides like {sup 169}Yb, bolus may be needed; and (iii) for the interstitial case, at least a 0.1 cm bolus is advised for {sup 60}Co to avoid underdosing superficial target layers. For {sup 192}Ir and {sup 169}Yb, no bolus is needed. For those cases where no bolus is needed, its use might be detrimental as the lack of radiation scatter may be beneficial to the patient, although the 2% tolerance for dose calculation accuracy recommended in the AAPM TG-56 report is not fulfilled.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2015

Measurement of natural radionuclides in Malaysian bottled mineral water and consequent health risk estimation

The radionuclides of {sup 226}Ra, {sup 232}Th and {sup 40}K were measured in ten mineral water samples, of which from the radioactivity obtained, the ingestion doses for infants, children and adults were calculated and the cancer risk for the adult was estimated. Results showed that the calculated ingestion doses for the three age categories are much lower than the average worldwide ingestion exposure of 0.29 mSv/y and the estimated cancer risk is much lower than the cancer risk of 8.40 × 10{sup −3} (estimated from the total natural radiation dose of 2.40 mSv/y). The present study concludes that the bottled mineral water produced in Malaysia is safe for daily human consumption.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2016

Radiological dosimetry measurements in Costa Rica

The main cause of human exposure to artificial radiation corresponds to medical applications, so it is essential to reduce the dose to patients, workers and consequently the entire population [1]. Although there is no dose limit for patients, is necessary to reduce it to a minimum possible while still getting all the necessary diagnostic information, taking economic and social factors into account [2]. Based on this proposal, agencies such as the International Atomic Energy Agency has been dedicated to providing guidelines levels, whose function is to serve as standards for the optimization of the medical exposure [3]. This research was created as a preliminary survey with the claim of eventually determine the guidance levels in Costa Rica for three different studies of general radiology: Lumbar Spine-AP, Chest - PA and Thoracic Spine - AP (for screens with speeds of 400 and 800), and cranio-caudal study in mammography, applied to Costa Rica’s adult population, perform properly in the institutions of Caja Costarricense del Seguro Social (CCSS).

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2015

SU-E-T-586: Optimal Determination of Tolerance Level for Radiation Dose Delivery Verification in An in Vivo Dosimetry System

Purpose: To statistically determine the optimal tolerance level in the verification of delivery dose compared to the planned dose in an in vivo dosimetry system in radiotherapy. Methods: The LANDAUER MicroSTARii dosimetry system with screened nanoDots (optically stimulated luminescence dosimeters) was used for in vivo dose measurements. Ideally, the measured dose should match with the planned dose and falls within a normal distribution. Any deviation from the normal distribution may be redeemed as a mismatch, therefore a potential sign of the dose misadministration. Randomly mis-positioned nanoDots can yield a continuum background distribution. A percentage difference of the measured dose to its corresponding planned dose (ΔD) can be used to analyze combined data sets for different patients. A model of a Gaussian plus a flat function was used to fit the ΔD distribution. Results: Total 434 nanoDot measurements for breast cancer patients were collected across a period of three months. The fit yields a Gaussian mean of 2.9% and a standard deviation (SD) of 5.3%. The observed shift of the mean from zero is attributed to the machine output bias and calibration of the dosimetry system. A pass interval of −2SD to +2SD was applied and a mismatch background was estimated to be 4.8%. With such a tolerance level, one can expect that 99.99% of patients should pass the verification and at most 0.011% might have a potential dose misadministration that may not be detected after 3 times of repeated measurements. After implementation, a number of new start breast cancer patients were monitored and the measured pass rate is consistent with the model prediction. Conclusion: It is feasible to implement an optimal tolerance level in order to maintain a low limit of potential dose misadministration while still to keep a relatively high pass rate in radiotherapy delivery verification.

61 RADIATION PROTECTION AND DOSIMETRY↗