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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

The difference of scoring dose to water or tissues in Monte Carlo dose calculations for low energy brachytherapy photon sources

Purpose: The goal of this work is to compare D{sub m,m} (radiation transported in medium; dose scored in medium) and D{sub w,m} (radiation transported in medium; dose scored in water) obtained from Monte Carlo (MC) simulations for a subset of human tissues of interest in low energy photon brachytherapy. Using low dose rate seeds and an electronic brachytherapy source (EBS), the authors quantify the large cavity theory conversion factors required. The authors also assess whether applying large cavity theory utilizing the sources' initial photon spectra and average photon energy induces errors related to spatial spectral variations. First, ideal spherical geometries were investigated, followed by clinical brachytherapy LDR seed implants for breast and prostate cancer patients. Methods: Two types of dose calculations are performed with the GEANT4 MC code. (1) For several human tissues, dose profiles are obtained in spherical geometries centered on four types of low energy brachytherapy sources: {sup 125}I, {sup 103}Pd, and {sup 131}Cs seeds, as well as an EBS operating at 50 kV. Ratios of D{sub w,m} over D{sub m,m} are evaluated in the 0-6 cm range. In addition to mean tissue composition, compositions corresponding to one standard deviation from the mean are also studied. (2) Four clinical breast (using {sup 103}Pd) and prostate (using {sup 125}I) brachytherapy seed implants are considered. MC dose calculations are performed based on postimplant CT scans using prostate and breast tissue compositions. PTV D{sub 90} values are compared for D{sub w,m} and D{sub m,m}. Results: (1) Differences (D{sub w,m}/D{sub m,m}-1) of -3% to 70% are observed for the investigated tissues. For a given tissue, D{sub w,m}/D{sub m,m} is similar for all sources within 4% and does not vary more than 2% with distance due to very moderate spectral shifts. Variations of tissue composition about the assumed mean composition influence the conversion factors up to 38%. (2) The ratio of D{sub 90(w,m)} over D{sub 90(m,m)} for clinical implants matches D{sub w,m}/D{sub m,m} at 1 cm from the single point sources. Conclusions: Given the small variation with distance, using conversion factors based on the emitted photon spectrum (or its mean energy) of a given source introduces minimal error. The large differences observed between scoring schemes underline the need for guidelines on choice of media for dose reporting. Providing such guidelines is beyond the scope of this work.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2014

SU-E-T-422: Correlation Between 2D Passing Rates and 3D Dose Differences for Pretreatment VMAT QA

Purpose: Volumetric modulated arc therapy (VMAT) quality assurance (QA) is typically using QA methods and action levels taken from fixedbeam intensity-modulated radiotherapy (IMRT) QA methods. However, recent studies demonstrated that there is no correlation between the percent gamma passing rate (%GP) and the magnitude of dose discrepancy between the planned dose and the actual delivered dose for IMRT. The purpose of this study is to investigate whether %GP is correlated with clinical dosimetric difference for VMAT. Methods: Twenty nasopharyngeal cancer (NPC) patients treated with dual-arc simultaneous integrated boost VMAT and 20 esophageal cancer patients treated with one-arc VMAT were enrolled in this study. Pretreatment VMAT QA was performed by a 3D diode array ArcCheck. Acceptance criteria of 2%/2mm, 3%/3mm, and 4%/4mm were applied for 2D %GP. Dose values below 10% of the per-measured normalization maximum dose were ignored.Mean DVH values obtained from 3DVH software and TPS were calculated and percentage dose differences were calculated. Statistical correlation between %GP and percent dose difference was studied by using Pearson correlation. Results: The %GP for criteria 2%/2mm, 3%/3mm, and 4%/4mm were 82.33±4.45, 93.47±2.31, 97.13±2.41, respectively. Dose differences calculated from 3DVH and TPS for beam isocenter, mean dose of PTV, maximum dose of PTV, D2 of PTV and D98 of PTV were -1.04±3.24, -0.74±1.71, 2.92±3.62, 0.89±3.29, -1.46±1.97, respectively. No correction were found between %GP and dose differences. Conclusion: There are weak correlations between the 2D %GP and dose differences calculated from 3DVH. The %GP acceptance criteria of 3%/3mm usually applied for pretreatment QA of IMRT and VMAT is not indicating strong clinical correlation with 3D dose difference. 3D dose reconstructions on patient anatomy may be necessary for physicist to predict the accuracy of delivered dose for VMAT QA.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

SU-F-T-477: Investigation of DEFGEL Dosimetry Using MRI

Purpose: The DEFGEL dosimeter/phantom allows for the measurement of 3D dose distributions while maintaining tissue equivalence and deformability. Although DEFGEL is traditionally read out with optical CT, the use of MRI would permit the measurement of 3D dose distributions in optically interfering configurations, like while embedded in a phantom. To the knowledge of the authors, this work is the first investigation that uses MRI to measure dose distributions in DEFGEL dosimeters. Methods: The DEFGEL (6%T) formula was used to create 1 cm thick, 4.5 cm diameter cylindrical dosimeters. The dosimeters were irradiated using a Varian Clinac 21EX linac. The MRI based transverse relaxation rate (R2) of the gel was measured in a central slice of the dosimeter with a Spin-Echo (SE) pulse sequence on a 3T GE SIGNA PET/MR scanner. The R2 values were fit to a monoexponential dose response equation using in-house software (MATLAB). Results: The data was well fit using a monoexponential fit for R2 as a function of absorbed dose (R{sup 2} = 0.9997). The fitting parameters of the monoexponential fit resulted in a 0.1229 Gy{sub −1}s{sub −1} slope. The data also resulted in an average standard deviation of 1.8% for the R2 values within the evaluated ROI. Conclusion: The close fit for the dose response curve shows that a DEFGEL based dosimeter can be paired with a SE MRI acquisition. The Type A uncertainty of the MRI method shows adequate precision, while the slope of the fit curve is large enough that R2 differences between different gel doses are distinguishable. These results suggest that the gel could potentially be used in configurations where an optical readout is not viable, such as measurements with the gel dosimeter positioned inside larger or optically opaque phantoms. This work is partially funded by NIH grant R01CA190298.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2015

Targeting mitochondria in cancer cells using gold nanoparticle-enhanced radiotherapy: A Monte Carlo study

Purpose: Radiation damage to mitochondria has been shown to alter cellular processes and even lead to apoptosis. Gold nanoparticles (AuNPs) may be used to enhance these effects in scenarios where they collect on the outer membranes of mitochondria. A Monte Carlo (MC) approach is used to estimate mitochondrial dose enhancement under a variety of conditions. Methods: The PENELOPE MC code was used to generate dose distributions resulting from photons striking a 13 nm diameter AuNP with various thicknesses of water-equivalent coatings. Similar dose distributions were generated with the AuNP replaced by water so as to estimate the gain in dose on a microscopic scale due to the presence of AuNPs within an irradiated volume. Models of mitochondria with AuNPs affixed to their outer membrane were then generated—considering variation in mitochondrial size and shape, number of affixed AuNPs, and AuNP coating thickness—and exposed (in a dose calculation sense) to source spectra ranging from 6 MV to 90 kVp. Subsequently dose enhancement ratios (DERs), or the dose with the AuNPs present to that for no AuNPs, for the entire mitochondrion and its components were tallied under these scenarios. Results: For a representative case of a 1000 nm diameter mitochondrion affixed with 565 AuNPs, each with a 13 nm thick coating, the mean DER over the whole organelle ranged from roughly 1.1 to 1.6 for the kilovoltage sources, but was generally less than 1.01 for the megavoltage sources. The outer membrane DERs remained less than 1.01 for the megavoltage sources, but rose to 2.3 for 90 kVp. The voxel maximum DER values were as high as 8.2 for the 90 kVp source and increased further when the particles clustered together. The DER exhibited dependence on the mitochondrion dimensions, number of AuNPs, and the AuNP coating thickness. Conclusions: Substantial dose enhancement directly to the mitochondria can be achieved under the conditions modeled. If the mitochondrion dose can be directly enhanced, as these simulations show, this work suggests the potential for both a tool to study the role of mitochondria in cellular response to radiation and a novel avenue for radiation therapy in that the mitochondria may be targeted, rather than the nuclear DNA.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2015

CIRCEE v1 - a versatile tool for criticality accident dose assessment - 14527

CIRCEE is a calculation code developed by the French Atomic Energy Commission (CEA), allowing fast assessment of operational dose over complex geometries. It is used to assess dosimetric consequences of a criticality accident. More specifically, it can be used both during the preliminary studies phase (isodoses curves for safety report preparation or emergency preparedness, CAAS implantation and validation...) and during crisis management (effective dose estimation, emergency response). An early development stage of CIRCEE has been presented in a talk at the 2009 NCSD edition [1]. Since then, an extensive code rewriting increased its reliability and maintainability while keeping its performance at an excellent level. Now in a stable production version, and following the grant of several international patents, it should be available under commercial license terms in the near future. (author)

42 ENGINEERING↗
NASA NTRS2018

Recent Developments in the Code RITRACKS (Relativistic Ion Tracks)

The code RITRACKS (Relativistic Ion Tracks) was developed to simulate detailed stochastic radiation track structures of ions of different types and energies. Many new capabilities were added to the code during the recent years. Several options were added to specify the times at which the tracks appear in the irradiated volume, allowing the simulation of dose-rate effects. The code has been used to simulate energy deposition in several targets: spherical, ellipsoidal and cylindrical. More recently, density changes as well as a spherical shell were implemented for spherical targets, in order to simulate energy deposition in walled tissue equivalent proportional counters. RITRACKS is used as a part of the new program BDSTracks (Biological Damage by Stochastic Tracks) to simulate several types of chromosome aberrations in various irradiation conditions. The simulation of damage to various DNA structures (linear and chromatin fiber) by direct and indirect effects has been improved and is ongoing. Many improvements were also made to the graphic user interface (GUI), including the addition of several labels allowing changes of units. A new GUI has been added to display the electron ejection vectors. The parallel calculation capabilities, notably the pre- and post-simulation processing on Windows and Linux machines have been reviewed to make them more portable between different systems. The calculation part is currently maintained in an Atlassian Stash® repository for code tracking and possibly future collaboration.

Plante, Ianik↗
DOE OSTI2013

Impact of Storage Time on the Needed Capture Efficiency for Volatile Radionuclides – 13369

During the processing of used nuclear fuel (UNF), volatile radionuclides will be discharged from the facility stack if no recovery processes are in place to limit their release. The volatile radionuclides of concern are 3 H, 14C, 85Kr, and 129I. There are three key regulations that address the release of these radionuclides to the environment — 40 CFR 61, 40 CFR 190, and 10 CFR 20. These regulations apply to the total radionuclide release and establish dose limits for the maximum exposed individual (MEI) in the public both in terms of whole body dose and dose to specific organs such as the thyroid. Each radionuclide released to the environment contributes to the total dose to some degree. In this paper we attempt to evaluate the efficiency requirements for the capture processes to limit the doses to the MEI to regulatory levels. Since the total amount of each volatile radionuclide present in the UNF changes with the age of the fuel, the respective capture requirements also change with time. Specifically, we are interested in the impact of the decreasing contribution of 3 H and 85Kr, which have relatively short half-lives, 12.32 y and 10.76 y, respectively, with the increasing age of the fuel (i.e., time between when the UNF is removed from the reactor and the time it is processed) on the capture requirements. In this paper we examine the capture requirements for these four radionuclides for three fuel types (pressurized water reactor [PWR] with uranium oxide fuel [UOX], PWR with mixed oxide fuel [MOX], and an advanced high temperature gas-cooled reactor [AHTGR]), several burnup values, and time out of reactor extending to 200 y. We calculate doses to the MEI with the EPA code CAP-88 and look at two dose contribution cases. In the first case, we assume that the total allowable dose is attributed to only the four volatile radionuclides. This establishes the lowest capture efficiency value possible. Since this is unrealistic, because it assumes zero dose contribution from all other radionuclides, we also examine a second case, where only 10% of the allowable dose is assigned to the four volatile radionuclides. We calculate the required decontamination factors (DFs) for both of these cases for the three fuel types, multiple fuel burnups, and fuel ages and determine whether or not the dose to the whole body or to the thyroid that drives the capture requirements would require additional effluent controls for the shorter half-life volatile radionuclides based on dose considerations. This analysis indicates that the principal isotopes of concern are generally 3 H and 129I, the latter requiring the highest DFs. The maximum DF value for 129I is 8000 for the evaluated cases and assumptions used.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Impact of Storage Time on the Needed Capture Efficiency for Volatile Radionuclides - 13369

During the processing of used nuclear fuel (UNF), volatile radionuclides will be discharged from the facility stack if no recovery processes are in place to limit their release. The volatile radionuclides of concern are {sup 3}H, {sup 14}C, {sup 85}Kr, and {sup 129}I. There are three key regulations that address the release of these radionuclides to the environment- 40 CFR 61, 40 CFR 190, and 10 CFR 20. These regulations apply to the total radionuclide release and establish dose limits for the maximum exposed individual (MEI) in the public both in terms of whole body dose and dose to specific organs such as the thyroid. Each radionuclide released to the environment contributes to the total dose to some degree. In this paper we attempt to evaluate the efficiency requirements for the capture processes to limit the doses to the MEI to regulatory levels. Since the total amount of each volatile radionuclide present in the UNF changes with the age of the fuel, the respective capture requirements also change with time. Specifically, we are interested in the impact of the decreasing contribution of {sup 3}H and {sup 85}Kr, which have relatively short half-lives, 12.32 y and 10.76 y, respectively, with the increasing age of the fuel (i.e., time between when the UNF is removed from the reactor and the time it is processed) on the capture requirements. In this paper we examine the capture requirements for these four radionuclides for three fuel types (pressurized water reactor [PWR] with uranium oxide fuel [UOX], PWR with mixed oxide fuel [MOX], and an advanced high temperature gas-cooled reactor [AHTGR]), several burnup values, and time out of reactor extending to 200 y. We calculate doses to the MEI with the EPA code CAP-88 and look at two dose contribution cases. In the first case, we assume that the total allowable dose is attributed to only the four volatile radionuclides. This establishes the lowest capture efficiency value possible. Since this is unrealistic, because it assumes zero dose contribution from all other radionuclides, we also examine a second case, where only 10% of the allowable dose is assigned to the four volatile radionuclides. We calculate the required decontamination factors (DFs) for both of these cases for the three fuel types, multiple fuel burnups, and fuel ages and determine whether or not the dose to the whole body or to the thyroid that drives the capture requirements would require additional effluent controls for the shorter half-life volatile radionuclides based on dose considerations. This analysis indicates that the principal isotopes of concern are generally {sup 3}H and {sup 129}I, the latter requiring the highest DFs. The maximum DF value for {sup 129}I is 8000 for the evaluated cases and assumptions used. ∼60 for fresh fuels. The DF for {sup 14}C could be as high as 30 for certain fuels. These values are based on just meeting the regulatory limits, and additional engineering margins (perhaps 3x to 10x or higher) should be applied to provide a safety factor for compliance. However, by assuming less conservative dose allocations, taller stacks, different radionuclide speciation, fuel aging, and other reprocessing facility design and location parameters, the DF requirements could be significantly reduced. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

An algorithm to calculate a collapsed arc dose matrix in volumetric modulated arc therapy

Purpose: The delivery of volumetric modulated arc therapy (VMAT) is more complex than other conformal radiotherapy techniques. In this work, the authors present the feasibility of performing routine verification of VMAT delivery using a dose matrix measured by a gantry mounted 2D ion chamber array and corresponding dose matrix calculated by an inhouse developed algorithm.Methods: Pinnacle, v9.0, treatment planning system (TPS) was used in this study to generate VMAT plans for a 6 MV photon beam from an Elekta-Synergy linear accelerator. An algorithm was developed and implemented with inhouse computer code to calculate the dose matrix resulting from a VMAT arc in a plane perpendicular to the beam at isocenter. The algorithm was validated using measurement of standard patterns and clinical VMAT plans with a 2D ion chamber array. The clinical VMAT plans were also validated using ArcCHECK measurements. The measured and calculated dose matrices were compared using gamma ({gamma}) analysis with 3%/3 mm criteria and {gamma} tolerance of 1.Results: The dose matrix comparison of standard patterns has shown excellent agreement with the mean {gamma} pass rate 97.7 ({sigma}= 0.4)%. The validation of clinical VMAT plans using the dose matrix predicted by the algorithm and the corresponding measured dose matrices also showed good agreement with the mean {gamma} pass rate of 97.6 ({sigma}= 1.6)%. The validation of clinical VMAT plans using ArcCHECK measurements showed a mean pass rate of 95.6 ({sigma}= 1.8)%.Conclusions: The developed algorithm was shown to accurately predict the dose matrix, in a plane perpendicular to the beam, by considering all possible leaf trajectories in a VMAT delivery. This enables the verification of VMAT delivery using a 2D array detector mounted on a treatment head.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

SU-F-T-316: A Model to Deal with Dosimetric and Delivery Uncertainties in Radiotherapy Treatment Planning

Purpose The conventional way of dealing with uncertainties resulting from dose calculation or beam delivery in IMRT, is to do verification measurements for the plan in question. Here we present an alternative based on recommendations given in the AAPM 142 report and treatment specific parameters that model the uncertainties for the plan delivery. Methods Basis of the model is the assignment of uncertainty parameters to all segment fields or control point sequences of a plan. The given field shape is analyzed for complexity, dose rate, number of MU, field size related output as well as factors for in/out field position and penumbra regions. Together with depth related uncertainties, a 3D matrix is generated by a projection algorithm. Patient anatomy is included as uncertainty CT data set as well. Therefore, object density is classified in 4 categories close to water, lung, bone and gradient regions with additional uncertainties. The result is then exported as a DICOM dose file by the software tool (written in IDL, Exelis), having the given resolution and target point. Results Uncertainty matrixes for several patient cases have been calculated and compared side by side in the planning system. The result is not quite always intuitive but it clearly indicates high and low uncertainties related to OARs and target volumes as well as to measured gamma distributions.ConclusionThe imported uncertainty datasets may help the treatment planner to understand the complexity of the treatment plan. He then might decide to change the plan to produce a more suited uncertainty distribution, e.g. by changing the beam angles the high uncertainty spots can be influenced or try to use another treatment setup, resulting in a plan with lower uncertainties. A next step could be to include such a model into the optimization algorithm to add a new dose uncertainty constraint.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2018

Development of ARCHER towards clinical use: modeling and simulation of Varian Linac for radiation therapy dose calculations - 25449

Monte Carlo particle transport methods are well recognized as the most accurate approach to dose calculations for radiation therapy treatment planning. However, the lengthy calculation time has prevented Monte Carlo methods from being applied to dose calculations in the planning of radiation therapy. Recently, the emerging parallel computing hardware accelerators, such as general-purpose Graphics Processing Unit (GPU), have significantly boosted the performance of Monte Carlo codes. The accuracy of a clinical dose engine depends greatly on the accuracy of source term modeling and the methodology of dose calculations. Hence, in this study, an efficient and accurate source model of the Varian TrueBeam{sup R} Linac is developed and integrated with a fast photon-electron transport module of the recently developed high-performance Monte Carlo code, ARCHER. The validity of the source model and the particle transport physics are verified with established dose calculation codes such as EGSnrc and validated with experimental measurements. To demonstrate the capability of the platform in clinical applications, three clinical treatment plans (the lung, breast, and prostate) are simulated and analyzed. From the perspective of performance, the entire simulation of each clinical plan, starting from the source modeling to the end of particle transport within the patient phantom, can be finished in less than one minute using a single Titan X GPU card with 1% statistical error. In comparison, the same cases take the EGSnrc code hundreds of CPU hours to finish. The results suggest that this new version of ARCHER code has been successfully developed and tested as a clinically deployable dose engine for external photon beam IMRT plans in the clinical setting. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

SU-F-T-46: The Effect of Inter-Seed Attenuation and Tissue Composition in Prostate 125I Brachytherapy Dose Calculations

Purpose: To investigate the difference of dose distributions with/without the effect of inter-seed attenuation and tissue compositions in prostate {sup 125}I brachytherapy dose calculations, using Monte Carlo simulations of Particle and Heavy Ion Transport code System (PHITS). Methods: The dose distributions in {sup 125}I prostate brachytherapy were calculated using PHITS for non-simultaneous and simultaneous alignments of STM1251 sources in water or prostate phantom for six patients. The PHITS input file was created from DICOM-RT file which includes source coordinates and structures for clinical target volume (CTV) and organs at risk (OARs) of urethra and rectum, using in-house Matlab software. Photon and electron cutoff energies were set to 1 keV and 100 MeV, respectively. The dose distributions were calculated with the kerma approximation and the voxel size of 1 × 1 × 1 mm{sup 3}. The number of incident photon was set to be the statistical uncertainty (1σ) of less than 1%. The effect of inter-seed attenuation and prostate tissue compositions was evaluated from dose volume histograms (DVHs) for each structure, by comparing to results of the AAPM TG-43 dose calculation (without the effect of inter-seed attenuation and prostate tissue compositions). Results: The dose reduction due to the inter-seed attenuation by source capsules was approximately 2% for CTV and OARs compared to those of TG-43. In additions, by considering prostate tissue composition, the D{sub 90} and V{sub 100} of CTV reduced by 6% and 1%, respectively. Conclusion: It needs to consider the dose reduction due to the inter-seed attenuation and tissue composition in prostate {sup 125}I brachytherapy dose calculations.

60 APPLIED LIFE SCIENCES↗
NASA NTRS2011

Total Ionizing Dose Testing of the Analog Devices’ AD648 Dual Precision Operational Amplifier

The Analog Devices (ADI) AD648 was tested for total ionizing dose response at the Goddard Space Flight Center (GSFC) Radiation Effects Facility (REF) beginning 1 June 2011. This test served as the radiation lot acceptance test (RLAT) for the lot date code (LDC) tested. Low dose rate (LDR) irradiations were performed in this test so that the device susceptibility to enhanced low dose rate sensitivity (ELDRS) could be determined.

Alvin J Boutte↗
DOE OSTI2014

SU-E-T-142: Effect of the Bone Heterogeneity On the Unflattened and Flattened Photon Beam Dosimetry: A Monte Carlo Comparison

Purpose: This study compared the dependence of depth dose on bone heterogeneity of unflattened photon beams to that of flattened beams. Monte Carlo simulations (the EGSnrc-based codes) were used to calculate depth doses in phantom with a bone layer in the buildup region of the 6 MV photon beams. Methods: Heterogeneous phantom containing a bone layer of 2 cm thick at a depth of 1 cm in water was irradiated by the unflattened and flattened 6 MV photon beams (field size = 10×10 cm{sup 2}). Phase-space files of the photon beams based on the Varian TrueBeam linac were generated by the Geant4 and BEAMnrc codes, and verified by measurements. Depth doses were calculated using the DOSXYZnrc code with beam angles set to 0° and 30°. For dosimetric comparison, the above simulations were repeated in a water phantom using the same beam geometry with the bone layer replaced by water. Results: Our results showed that the beam output of unflattened photon beams was about 2.1 times larger than the flattened beams in water. Comparing the water phantom to the bone phantom, larger doses were found in water above and below the bone layer for both the unflattened and flattened photon beams. When both beams were turned 30°, the deviation of depth dose between the bone and water phantom became larger compared to that with beam angle equal to 0°. Dose ratio of the unflattened and flattened photon beams showed that the unflattened beam has larger depth dose in the buildup region compared to the flattened beam. Conclusion: Although the unflattened photon beam had different beam output and quality compared to the flattened, dose enhancements due to the bone scatter were found similar. However, we discovered that depth dose deviation due to the presence of bone was sensitive to the beam obliquity.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2011

Development of a patient-specific dosimetry estimation system in nuclear medicine examination

The purpose of this study is to develop a patient-specific dosimetry estimation system in nuclear medicine examination using a SimSET-based Monte Carlo code. We added a dose deposition routine to store the deposited energy of the photons during their flights in SimSET and developed a user-friendly interface for reading PET and CT images. Dose calculated on ORNL phantom was used to validate the accuracy of this system. The S values for {sup 99m}Tc, {sup 18}F and {sup 131}I obtained by the system were compared to those from the MCNP4C code and OLINDA. The ratios of S values computed by this system to those obtained with OLINDA for various organs were ranged from 0.93 to 1.18, which are comparable to that obtained from MCNP4C code (0.94 to 1.20). The average ratios of S value were 0.99{+-}0.04, 1.03{+-}0.05, and 1.00{+-}0.07 for isotopes {sup 131}I, {sup 18}F, and {sup 99m}Tc, respectively. The simulation time of SimSET was two times faster than MCNP4C's for various isotopes. A 3D dose calculation was also performed on a patient data set with PET/CT examination using this system. Results from the patient data showed that the estimated S values using this system differed slightly from those of OLINDA for ORNL phantom. In conclusion, this system can generate patient-specific dose distribution and display the isodose curves on top of the anatomic structure through a friendly graphic user interface. It may also provide a useful tool to establish an appropriate dose-reduction strategy to patients in nuclear medicine environments. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

SU-E-I-63: Quantitative Evaluation of the Effects of Orthopedic Metal Artifact Reduction (OMAR) Software On CT Images for Radiotherapy Simulation

Purpose: CT simulation for patients with metal implants can often be challenging due to artifacts that obscure tumor/target delineation and normal organ definition. Our objective was to evaluate the effectiveness of Orthopedic Metal Artifact Reduction (OMAR), a commercially available software, in reducing metal-induced artifacts and its effect on computed dose during treatment planning. Methods: CT images of water surrounding metallic cylindrical rods made of aluminum, copper and iron were studied in terms of Hounsfield Units (HU) spread. Metal-induced artifacts were characterized in terms of HU/Volume Histogram (HVH) using the Pinnacle treatment planning system. Effects of OMAR on enhancing our ability to delineate organs on CT and subsequent dose computation were examined in nine (9) patients with hip implants and two (2) patients with breast tissue expanders. Results: Our study characterized water at 1000 HU with a standard deviation (SD) of about 20 HU. The HVHs allowed us to evaluate how the presence of metal changed the HU spread. For example, introducing a 2.54 cm diameter copper rod in water increased the SD in HU of the surrounding water from 20 to 209, representing an increase in artifacts. Subsequent use of OMAR brought the SD down to 78. Aluminum produced least artifacts whereas Iron showed largest amount of artifacts. In general, an increase in kVp and mA during CT scanning showed better effectiveness of OMAR in reducing artifacts. Our dose analysis showed that some isodose contours shifted by several mm with OMAR but infrequently and were nonsignificant in planning process. Computed volumes of various dose levels showed <2% change. Conclusions: In our experience, OMAR software greatly reduced the metal-induced CT artifacts for the majority of patients with implants, thereby improving our ability to delineate tumor and surrounding organs. OMAR had a clinically negligible effect on computed dose within tissues. Partially funded by unrestricted educational grant from Philips.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

Reoptimization of Intensity Modulated Proton Therapy Plans Based on Linear Energy Transfer

Purpose: We describe a treatment plan optimization method for intensity modulated proton therapy (IMPT) that avoids high values of linear energy transfer (LET) in critical structures located within or near the target volume while limiting degradation of the best possible physical dose distribution. Methods and Materials: To allow fast optimization based on dose and LET, a GPU-based Monte Carlo code was extended to provide dose-averaged LET in addition to dose for all pencil beams. After optimizing an initial IMPT plan based on physical dose, a prioritized optimization scheme is used to modify the LET distribution while constraining the physical dose objectives to values close to the initial plan. The LET optimization step is performed based on objective functions evaluated for the product of LET and physical dose (LET×D). To first approximation, LET×D represents a measure of the additional biological dose that is caused by high LET. Results: The method is effective for treatments where serial critical structures with maximum dose constraints are located within or near the target. We report on 5 patients with intracranial tumors (high-grade meningiomas, base-of-skull chordomas, ependymomas) in whom the target volume overlaps with the brainstem and optic structures. In all cases, high LET×D in critical structures could be avoided while minimally compromising physical dose planning objectives. Conclusion: LET-based reoptimization of IMPT plans represents a pragmatic approach to bridge the gap between purely physical dose-based and relative biological effectiveness (RBE)-based planning. The method makes IMPT treatments safer by mitigating a potentially increased risk of side effects resulting from elevated RBE of proton beams near the end of range.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2025

Dose Exceedance Distance Sensitivity Based on Parametric Uncertainty

Sandia National Laboratories (SNL) collaborated with the US Nuclear Regulatory Commission's (NRC) Office of Nuclear Security and Incident Response (NSIR) and Office of Nuclear Regulatory Research (RES) to investigate the reasonable variability in the estimation of dose exceedance distances. SNL performed calculations using the MACCS code to elucidate the sensitivity of dose exceedance distances to variations in user input parameters.

61 RADIATION PROTECTION AND DOSIMETRY↗