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DOE OSTI2012

DOE 2012 Occupational Radiation Exposure October 2013

The U.S. Department of Energy (DOE) Office of Analysis within the Office of Health, Safety and Security (HSS) publishes the annual DOE Occupational Radiation Exposure Report to provide an overview of the status of radiation protection practices at DOE (including the National Nuclear Security Administration [NNSA]). The DOE 2012 Occupational Radiation Exposure Report provides an evaluation of DOE-wide performance regarding compliance with Title 10, Code of Federal Regulations (C.F.R.), Part 835, Occupational Radiation Protection dose limits and as low as reasonably achievable (ALARA) process requirements. In addition, the report provides data to DOE organizations responsible for developing policies for protection of individuals from the adverse health effects of radiation. The report provides a summary and an analysis of occupational radiation exposure information from the monitoring of individuals involved in DOE activities. Over the past 5-year period, the occupational radiation exposure information is analyzed in terms of aggregate data, dose to individuals, and dose by site. As an indicator of the overall amount of radiation dose received during the conduct of operations at DOE, the report includes information on collective total effective dose (TED). The TED is comprised of the effective dose (ED) from external sources, which includes neutron and photon radiation, and the internal committed effective dose (CED), which results from the intake of radioactive material into the body. The collective ED from photon exposure decreased by 23% between 2011 and 2012, while the neutron dose increased by 5%. The internal dose components of the collective TED decreased by 7%. Over the past 5-year period, 99.99% of the individuals receiving measurable TED have received doses below the 2 roentgen equivalent in man (rems) (20 millisievert [mSv]) TED administrative control level (ACL), which is well below the DOE regulatory limit of 5 rems (50 mSv) TED annually. The occupational radiation exposure records show that in 2012, DOE facilities continued to comply with DOE dose limits and ACLs and worked to minimize exposure to individuals. The DOE collective TED decreased 17.1% from 2011 to 2012. The collective TED decreased at three of the five sites with the largest collective TED. u Idaho Site – Collective dose reductions were achieved as a result of continuing improvements at the Advanced Mixed Waste Treatment Project (AMWTP) through the planning of drum movements that reduced the number of times a container is handled; placement of waste containers that created highradiation areas in a centralized location; and increased worker awareness of high-dose rate areas. In addition, Idaho had the largest decrease in the total number of workers with measurable TED (1,143 fewer workers). u Hanford Site (Hanford) – An overall reduction of decontamination and decommissioning (D&D) activities at the Plutonium Finishing Plant (PFP) and Transuranic (TRU) retrieval activities resulted in collective dose reductions. u Savannah River Site (SRS) – Reductions were achieved through ALARA initiatives employed site wide. The Solid Waste Management Facility used extended specialty tools, cameras and lead shield walls to facilitate removal of drums. These tools and techniques reduce exposure time through improved efficiency, increase distance from the source of radiation by remote monitoring, shield the workers to lower the dose rate, and reduce the potential for contamination and release of material through repacking of waste. Overall, from 2011 to 2012, there was a 19% decrease in the number of workers with measurable dose. Furthermore, due to a slight decrease in both the DOE workforce (7%) and monitored workers (10%), the ratio of workers with measurable doses to monitored workers decreased to 13%. Another primary indicator of the level of radiation exposure covered in this report is the average measurable dose, which normalizes the collective dose over the population of workers who actually received a measurable dose. The average measurable TED increased by 3% from 2011 to 2012. Additional analyses show that the dose distribution in 2012 was similar to the distribution in 2011. In 2012, 13% of the monitored workers received a measurable TED and the average measurable TED, 0.069 rem, was less than 2% of the DOE limit. From 2011 to 2012, the collective TED and the number of individuals with measurable TED decreased 17.1% and 19%, respectively. These decreases were mainly due to an overall reduction of D&D activities at the PFP and TRU retrieval activities at Hanford; a 78% decrease in the number of targeted waste drums that were processed at the Idaho Site’s Accelerated Retrieval Project (ARP) from 5,566 drums in 2011 to a total of 1,211 drums processed in 2012; and ALARA initiatives employed site wide at SRS. In addition, the decreases were the result of decreased American Recovery and Reinvestment Act (ARRA) activities and continuing D&D, particularly at the DOE sites that comprise the majority of DOE collective dose. Over the past 5 years, the size of the monitored workforce has remained at a fairly stable level (within 12%), while the collective dose has varied up to 37%. No reported doses exceeded the DOE occupational limit of 5 rems TED in 2012 and no reported doses exceeded the DOE ACL of 2 rems TED.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2014

DOE 2013 occupational radiation exposure

The Office of Analysis within the U.S. Department of Energy (DOE) Office of Environment, Health, Safety and Security (EHSS) publishes the annual DOE Occupational Radiation Exposure Report to provide an overview of the status of radiation protection practices at DOE (including the National Nuclear Security Administration [NNSA]). The DOE 2013 Occupational Radiation Exposure Report provides an evaluation of DOE-wide performance regarding compliance with Title 10, Code of Federal Regulations (C.F.R.), Part 835, Occupational Radiation Protection dose limits and as low as reasonably achievable (ALARA) process requirements. In addition, the report provides data to DOE organizations responsible for developing policies for protection of individuals from the adverse health effects of radiation. The report provides a summary and an analysis of occupational radiation exposure information from the monitoring of individuals involved in DOE activities. Over the past five-year period, the occupational radiation exposure information has been analyzed in terms of aggregate data, dose to individuals, and dose by site.

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DOE OSTI2013

DOE 2012 occupational radiation exposure

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DOE OSTI2012

DOE 2011 occupational radiation exposure

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DOE OSTI2021

Evaluation of Attila and MCNP computational methods for dose and exposure estimation

Radiation transport calculations are often used to estimate dose or exposure to components and personnel surrounding a radiation source. The sources for these calculations are decaying radionuclides within various nuclear materials. Historically, dose calculations use MCNP (Monte Carlo N-Particle) transport code as the primary particle transport tool without a secondary computational tool to validate the results from the MCNP simulations [1]. The goal of this study is to make an independent check of the Monte Carlo solution from MCNP6 Version 6.2.1 with the discrete ordinates solution from Attila 10.2.0 Beta 3. As an example problem for this study, water-filled, stainless-steel vessels, modeled with an unstructured mesh (UM) with both MCNP and Attila [2], are exposed to 252Cf and 60Co point sources. This report also includes a discussion of the limitations of unstructured mesh in a MCNP calculation.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2024

Activation estimates of the RHIC beam dump

In anticipation of the expiration of the RHIC ASE in December 2025 and the need to demonstrate compliance with nuclear facility rule (10CFR830), this study quantifies the radionuclide inventory and respective activity at the beam dumps by simulating a period of five years of operation using conservative assumptions. Simulations were performed with the Monte Carlo particle transport and interaction code FLUKA and its dosimetry results were compared with measured dose rate data, showing good agreement. The beam dump was deemed to constitute the main hotspot of residual radiation and also representative of a worst case scenario in terms of induced activation. The activation levels calculated indicate that the RHIC tunnel will be well below the thresholds of a DOE’s category-3 nuclear facility classification.

43 PARTICLE ACCELERATORS↗

NASA NTRS1988

Solar-flare shielding with Regolith at a lunar-base site

The Langley high energy nucleon transport computer code BRYNTRN is used to predict time-integrated radiation dose levels at the lunar surface due to high proton flux from solar flares. The study addresses the shielding requirements for candidate lunar habitat configurations necessary to protect crew members from these large and unpredictable radiation fluxes. Three solar proton events have been analyzed, and variations in radiation intensity in a shield medium due to the various primary particle energy distributions are predicted. Radiation dose predictions are made for various slab thicknesses of a lunar soil model. Results are also presented in the form of dose patterns within specific habitat configurations shielded with lunar material.

Nealy, John E.↗

NASA NTRS2005

The Application of FLUKA to Dosimetry and Radiation Therapy

Monte Carlo transport codes like FLUKA are useful for many purposes, and one of those is the simulation of the effects of radiation traversing the human body. In particular, radiation has been used in cancer therapy for a long time, and recently this has been extended to include heavy ion particle beams. The advent of this particular type of therapy has led to the need for increased capabilities in the transport codes used to simulate the detailed nature of the treatment doses to the Y ~ ~ O U S tissues that are encountered. This capability is also of interest to NASA because of the nature of the radiation environment in space.[l] While in space, the crew members bodies are continually being traversed by virtually all forms of radiation. In assessing the risk that this exposure causes, heavy ions are of primary importance. These arise both from the primary external space radiation itself, as well as fragments that result from interactions during the traversal of that radiation through any intervening material including intervening body tissue itself. Thus the capability to characterize the details of the radiation field accurately within a human body subjected to such external 'beams" is of critical importance.

Wilson, Thomas L.↗

NASA NTRS2017

3DHZETRN: Inhomogeneous Geometry Issues

Historical methods for assessing radiation exposure inside complicated geometries for space applications were limited by computational constraints and lack of knowledge associated with nuclear processes occurring over a broad range of particles and energies. Various methods were developed and utilized to simplify geometric representations and enable coupling with simplified but efficient particle transport codes. Recent transport code development efforts, leading to 3DHZETRN, now enable such approximate methods to be carefully assessed to determine if past exposure analyses and validation efforts based on those approximate methods need to be revisited. In this work, historical methods of representing inhomogeneous spacecraft geometry for radiation protection analysis are first reviewed. Two inhomogeneous geometry cases, previously studied with 3DHZETRN and Monte Carlo codes, are considered with various levels of geometric approximation. Fluence, dose, and dose equivalent values are computed in all cases and compared. It is found that although these historical geometry approximations can induce large errors in neutron fluences up to 100 MeV, errors on dose and dose equivalent are modest (<10%) for the cases studied here.

Wilson, John W.↗

DOE OSTI2014

SU-E-J-72: Geant4 Simulations of Spot-Scanned Proton Beam Treatment Plans

Purpose: To evaluate density inhomogeneities which can effect dose distributions for real-time image gated spot-scanning proton therapy (RGPT), a dose calculation system, using treatment planning system VQA (Hitachi Ltd., Tokyo) spot position data, was developed based on Geant4. Methods: A Geant4 application was developed to simulate spot-scanned proton beams at Hokkaido University Hospital. A CT scan (0.98 × 0.98 × 1.25 mm) was performed for prostate cancer treatment with three or four inserted gold markers (diameter 1.5 mm, volume 1.77 mm3) in or near the target tumor. The CT data was read into VQA. A spot scanning plan was generated and exported to text files, specifying the beam energy and position of each spot. The text files were converted and read into our Geant4-based software. The spot position was converted into steering magnet field strength (in Tesla) for our beam nozzle. Individual protons were tracked from the vacuum chamber, through the helium chamber, steering magnets, dose monitors, etc., in a straight, horizontal line. The patient CT data was converted into materials with variable density and placed in a parametrized volume at the isocenter. Gold fiducial markers were represented in the CT data by two adjacent voxels (volume 2.38 mm3). 600,000 proton histories were tracked for each target spot. As one beam contained about 1,000 spots, approximately 600 million histories were recorded for each beam on a blade server. Two plans were considered: two beam horizontal opposed (90 and 270 degree) and three beam (0, 90 and 270 degree). Results: We are able to convert spot scanning plans from VQA and simulate them with our Geant4-based code. Our system can be used to evaluate the effect of dose reduction caused by gold markers used for RGPT. Conclusion: Our Geant4 application is able to calculate dose distributions for spot scanned proton therapy.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2015

SU-E-T-212: Comparison of TG-43 Dosimetric Parameters of Low and High Energy Brachytherapy Sources Obtained by MCNP Code Versions of 4C, X and 5

Purpose: Different versions of MCNP code are widely used for dosimetry purposes. The purpose of this study is to compare different versions of the MCNP codes in dosimetric evaluation of different brachytherapy sources. Methods: The TG-43 parameters such as dose rate constant, radial dose function, and anisotropy function of different brachytherapy sources, i.e. Pd-103, I-125, Ir-192, and Cs-137 were calculated in water phantom. The results obtained by three versions of Monte Carlo codes (MCNP4C, MCNPX, MCNP5) were compared for low and high energy brachytherapy sources. Then the cross section library of MCNP4C code was changed to ENDF/B-VI release 8 which is used in MCNP5 and MCNPX codes. Finally, the TG-43 parameters obtained using the MCNP4C-revised code, were compared with other codes. Results: The results of these investigations indicate that for high energy sources, the differences in TG-43 parameters between the codes are less than 1% for Ir-192 and less than 0.5% for Cs-137. However for low energy sources like I-125 and Pd-103, large discrepancies are observed in the g(r) values obtained by MCNP4C and the two other codes. The differences between g(r) values calculated using MCNP4C and MCNP5 at the distance of 6cm were found to be about 17% and 28% for I-125 and Pd-103 respectively. The results obtained with MCNP4C-revised and MCNPX were similar. However, the maximum difference between the results obtained with the MCNP5 and MCNP4C-revised codes was 2% at 6cm. Conclusion: The results indicate that using MCNP4C code for dosimetry of low energy brachytherapy sources can cause large errors in the results. Therefore it is recommended not to use this code for low energy sources, unless its cross section library is changed. Since the results obtained with MCNP4C-revised and MCNPX were similar, it is concluded that the difference between MCNP4C and MCNPX is their cross section libraries.

60 APPLIED LIFE SCIENCES↗

NASA NTRS2014

Implementing Badhwar-O'Neill Galactic Cosmic Ray Model for the Analysis of Space Radiation Exposure

For the analysis of radiation risks to astronauts and planning exploratory space missions, accurate energy spectrum of galactic cosmic radiation (GCR) is necessary. Characterization of the ionizing radiation environment is challenging because the interplanetary plasma and radiation fields are modulated by solar disturbances and the radiation doses received by astronauts in interplanetary space are likewise influenced. A model of the Badhwar‐O'Neill 2011 (BO11) GCR environment, which is represented by GCR deceleration potential theta, has been derived by utilizing all of the GCR measurements from balloons, satellites, and the newer NASA Advanced Composition Explorer (ACE). In the BO11 model, the solar modulation level is derived from the mean international sunspot numbers with time‐delay, which has been calibrated with actual flight instrument measurements to produce better GCR flux data fit during solar minima. GCR fluxes provided by the BO11 model were compared with various spacecraft measurements at 1 AU, and further comparisons were made for the tissue equivalent proportional counters measurements at low Earth orbits using the high‐charge and energy transport (HZETRN) code and various GCR models. For the comparison of the absorbed dose and dose equivalent calculations with the measurements by Radiation Assessment Detector (RAD) at Gale crater on Mars, the intensities and energies of GCR entering the heliosphere were calculated by using the BO11 model, which accounts for time‐dependent attenuation of the local interstellar spectrum of each element. The BO11 model, which has emphasized for the last 24 solar minima, showed in relatively good agreement with the RAD data for the first 200 sols, but it was resulted in to be less well during near the solar maximum of solar cycle 24 due to subtleties in the changing heliospheric conditions. By performing the error analysis of the BO11 model and the optimization in reducing overall uncertainty, the resultant BO13 model corrects the fit at solar maxima as well as being accurate at solar minima. The BO13 model is implemented to the NASA Space Cancer Risk model for the assessment of radiation risks. Overall cumulative probability distribution of solar modulation parameters represents the percentile rank of the average interplanetary GCR environment, and the probabilistic radiation risks can be assessed for various levels of GCR environment to support mission design and operational planning for future manned space exploration missions.

Kim, Myung-Hee Y.↗

DOE OSTI2020

The Relationship Between Dose Rate and Decay Heat for Spent Nuclear Fuel Casks

Decay heat and dose rate are two important limits used for determining the allowable contents of spent fuel (SNF) in dry storage systems and transportation packages. While the decay heat limit is used to maintain fuel cladding integrity and ensure retrievability, dose rates are used to demonstrate compliance with regulatory requirements on radiation protection. Because both dose rate and decay heat result from decay of radioisotopes in SNF, this study is an attempt to examine the relationship between dose rate and decay heat for a given cask design. Dose rates were evaluated for 198 cask configurations, that include various SNF system designs (e.g., storage, transfer, transport), SNF characteristics (e.g., fuel types, burnup, cooling time), and loading maps (e.g., uniform loading, zone loading), while a constant decay heat was maintained. The decay heat was calculated using US Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 3.54, Revision 2, and verified using ORIGEN sequence within SCALE code system. The ORIGEN outputs were used as source terms in the dose analysis using 198 different configurations. A computer script was developed to calculate the cooling time necessary to achieve a given decay heat for a given enrichment, assembly average burnup, assembly mass, and in-core history using a rootfinder algorithm. Initially the computer script was developed to provide cooling time and burnup calculations directly to the analysis of dose rates and decay heats, so a comparison between Used Nuclear Fuel-Storage Transportation and Disposal Analysis Resource Data System (UNF-ST&DARDS) results and RG3.54 data was made; results are included in the appendix to this document. However, an iterative approach was used to compute cooling time, and its accuracy did not depend on the results of the RG3.54r2 algorithm, although the algorithm was still used. The results of the evaluation presented herein clearly demonstrate that a given decay heat does not correspond to a unique dose rate for a variety of cask and package designs. There is no clear pattern to develop a correlation between decay heat and the source terms. Depending on burnup, enrichment, cask type, and loading pattern, dose rates varied for the exact same decay heat—in some cases by 400% for a given cask. For cases in which decay heat was held constant through selection of the appropriate cooling time, dose rates would decrease with increasing burnup, and in other cases, dose rates would increase. The large variation in dose rates for a constant decay heat indicates that casks loaded based on decay heat—that is allowing any burnup, cooling time, and enrichment combinations that yield the qualified decay heat limit(s) —cannot ensure that an Independent Spent Fuel Storage Installation or a spent fuel transportation package will meet the regulatory limits set forth by the respective regulations, i.e., 10 CFR 72 or 10 CFR 71.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2016

SU-G-BRC-16: Theory and Clinical Implications of the Constant Dosimetric Leaf Gap (DLG) Approximation

Purpose: Commercial dose calculation algorithms incorporate a single DLG value for a given beam energy that is applied across an entire treatment field. However, the physical processes associated with beam generation and dose delivery suggest that the DLG is not constant. The aim of this study is to evaluate the variation of DLG among all leaf pairs, to quantify how this variation impacts delivered dose, and to establish a novel method to correct dose distributions calculated using the approximation of constant DLG. Methods: A 2D diode array was used to measure the DLG for all 60 leaf pairs at several points along each leaf pair travel direction. This approach was validated by comparison to DLG values measured at select points using a 0.6 cc ion chamber with the standard formalism. In-house software was developed to enable incorporation of position dependent DLG values into dose distribution optimization and calculation. The accuracy of beam delivery of both the corrected and uncorrected treatment plans was studied through gamma pass rate evaluation. A comparison of DVH statistics in corrected and uncorrected treatment plans was made. Results: The outer 20 MLC leaf pairs (1.0 cm width) have DLG values that are 0.32 mm (mean) to 0.65 mm (maximum) lower than the central leaf-pair. VMAT plans using a large number of 1 cm wide leaves were more accurately delivered (gamma pass rate increased by 5%) and dose coverage was higher (D100 increased by 3%) when the 2D DLG was modeled. Conclusion: Using a constant DLG value for a given beam energy will result in dose optimization, dose calculation and treatment delivery inaccuracies that become significant for treatment plans with high modulation complexity scores delivered with 1 cm wide leaves.

60 APPLIED LIFE SCIENCES↗

DOE OSTI2025

Status of the MARS code

This report describes major features of the most recent version of the MARS code as well as ongoing developments. The list of features includes various options for geometry models, a beam line builder based on MADX code, import of geometry models in GDML format, use of structured and unstructured meshes for scoring purposes, an update to the recent TENDL library for a number of projectiles at low energies (up to 250 MeV), and a recently implemented method to calculate spatial distribution of residual dose in a single computer run without an intermediate source. Examples of the code application to various projects are presented as well.

Rakhno, Igor [Fermilab] (ORCID:0000000265828058)↗

DOE OSTI2010

GEANT4 Application for the Simulation of the Head of a Siemens Primus Linac

The Monte Carlo simulation of the head of a Siemens Primus Linac used at Virgen Macarena Hospital (Sevilla, Spain) has been performed using the code GEANT4, version 9.2. In this work, the main features of the application built by our group are presented. They are mainly focused in the optimization of the performance of the simulation. The geometry, including the water phantom, has been entirely wrapped by a shielding volume which discards all the particles escaping far away through its walls. With this, a factor of four in the time spent by the simulation can be saved. An interface to read and write phase-space files in IAEA format has been also developed to save CPU time in our simulations. Finally, some calculations of the dose absorption in the water phantom have been done and compared with the results given by EGSnrc and with experimental data obtained for the calibration of the machine.

43 PARTICLE ACCELERATORS↗

DOE OSTI2014

WE-A-17A-07: Evaluation of a Grid-Based Boltzmann Solver for Nuclear Medicine Voxel-Based Dose Calculations

Purpose: Grid-based Boltzmann solvers (GBBS) have been successfully implemented in radiation oncology clinics as dose calculations for e×ternal photon beams and 192Ir sealed-source brachytherapy. We report on the evaluation of a GBBS for nuclear medicine vo×el-based absorbed doses. Methods: Vo×el-S-values were calculated for monoenergetic betas and photons (1, 0.1, 0.01 MeV), 90Y, and 131I for 3 mm vo×el sizes using Monte Carlo (DOS×YZnrc) and GBBS (Attila 8.1-beta5, Transpire). The source distribution was uniform throughout a single vo×el. The material was an infinite 1.04 g/cc soft tissue slab. To e×plore convergence properties of the GBBS 3 tetrahedral meshes, 3 energy group structures, 3 different square Chebyschev-Legendre quadrature set orders (Sn), and 4×2013;7 spherical harmonic e×pansion terms (Pn) were investigated for a total of 168 discretizations per source. The mesh, energy group, and quadrature sets are 8×, 3×, and 16×, respectively, finer than the corresponding coarse discretization. GBBS cross sections were generated with full electronphoton-coupling using the vendors e×tended CEP×S code. For accuracy, percent differences (%Δ) in source vo×el absorbed doses between MC and GBBS are reported for the coarsest and finest discretization. For convergence, ratios of the two finest discretization solutions are reported along each variable. Results: For 1 MeV, 0.1 MeV, 0.01 MeV, Y90, and I-131 beta sources the %Δ in the source vo×el for (coarsest,finest) discretization were (+2.0,−6.4), (−8.0, −7.5), (−13.8, −13.4), (+0.9,−5.5), and (− 10.1,−9.0) respectively. The corresponding %Δ for photons were (+33.7,−7.1), (−9.4, −9.8), (−17.4, −15.2), and (−1.7,−7.7), respectively. For betas, the convergence ratio of mesh, energy, Sn, and Pn ranged from 0.991–1.000. For gammas, the convergence ratio of mesh, Sn, and Pn ranged from 0.998–1.003 while the ratio for energy ranged from 0.964–1.001. Conclusions: GBBS is promising for nuclear medicine vo×el-based dose calculations. Ongoing work includes evaluating GBBS in bone, lung, and realistic clinical PET/SPECT-based activity distributions. Research reported in this publication was supported by the National Cancer Institute of the National Institutes of Health under Award Number R01CA138986. The content is solely the responsibility of the authors and does not necessarily represent the official views of the National Institutes of Health.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2016

SU-F-T-137: Out-Of-Beam Dose for a Compact Double-Scattering Proton Beam Therapy System

Purpose: The out-of-beam dose is important for understanding the peripheral dose in radiation therapy. In proton radiotherapy, the study of out-of-beam dose is scarce and the treatment planning system (TPS) based on pencil beam algorithm cannot accurately predict the out-of-beam dose. This study investigates the out-of-beam dose for the single-room Mevion S250 double scattering proton therapy system using experimentally measured and treatment planning software generated data. The results are compared with those reported for conventional photon beam therapy. However, this study does not incorporate the neutron contribution in the scattered dose. Methods: A total of seven proton treatment plans were generated using Varian Eclipse TPS for three different sites (brain, lung, and pelvis) in an anthropomorphic phantom. Three field sizes of 5×5, 10×10, and 20×20 cm{sup 2} (lung only) with typical clinical range (13.3–22.8 g/cm{sup 2}) and modulation widths (5.3–14.0 g/cm{sup 2}) were used. A single beam was employed in each treatment plan to deliver a dose of 181.8 cGy (200.0 cGy (RBE)) to the selected target. The out-of-beam dose was measured at 2.0, 5.0, 10.0, and 15.0 cm from the beam edge in the phantom using a thimble chamber (PTW TN31010). Results: The out-of-beam dose generally increased with field size, range, and volume irradiated. For all the plans, the scattered dose sharply fell off with distance. At 2.0 cm, the out-of-beam dose ranged from 0.35% to 2.16% of the delivered dose; however, the dose was clinically negligible (<0.3%) at a distance of 5.0 cm and greater. In photon therapy, the slightly greater out-of-beam dose was reported (TG36; 4%, 2%, and 1% for 2.0, 5.0, and 10.0 cm, respectively, using 6 MV beam). Conclusion: The measured out-of-beam dose in proton therapy excluding neutron contribution was observed higher than the TPS calculated dose and comparable to that of photon beam therapy.

60 APPLIED LIFE SCIENCES↗