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At least 1,243 records · Page 69

Dose Rate Investigation for Cobalt Isotope Rods Retrieval Campaign

GE-Hitachi (GEH) partnered with Exelon for a first of a kind program to generate cobalt-60 (Co-60) sources at Clinton Power Station (CPS). The cobalt targets are placed within the fuel bundle, double encapsulated in metal rods, and irradiated in the Boiling Water Reactor for activation. After the bundle is discharged, the Co-60 sources are harvested for medical and industrial applications. In order to ensure safe shipment of the Co-60 sources, shielding evaluations are required. As required by 10 CFR 71, a licensing basis evaluation was performed and submitted to the Nuclear Regulatory Commission (NRC) and approved as a Letter Authorization. This 10 CFR 71 evaluation utilizes bounding design inputs to demonstrate compliance with the regulatory limits. The retrieval campaign of the Co-60 rods involves disassembly of the bundle/rods, loading into a material basket, and then loading the material basket into GEHitachi's Model 2000 cask that is part of the Type B Transportation Package. As part of conservative decision making, CPS wanted confidence that measurements taken at various locations around the material basket loaded with Co-60 rods are consistent or bounded by the dose rate calculations performed by GEH. This would ensure that after all loading operations are completed, and the Model 2000 package is in the shipping configuration (i.e., cask drained and secured within the Model 2000 overpack), the dose rates would meet 10 CFR 71 shipping requirements. In addition, dose rate predictions were used to develop an ALARA plan at CPS to ensure radiological safety of the workers. To support CPS, a second, more realistic shielding evaluation was performed to verify the as planned loading. (authors)

07 ISOTOPE AND RADIATION SOURCES↗

TENDL-2017: the making of multi-faceted technological nuclear data library - 25340

Multi-purpose methodical nuclear data libraries generated by TALYS/T6 as Evaluated Nuclear Data Library TENDL have been released regularly since 2008. Considerable experience has been acquired during those early production of such unique, truly multi-faceted, multi-particle nuclear data libraries based on the feedback from the developers, evaluators, processing experts and, most importantly, users. The backbone of this achievement is simple and robust: completeness, quality, upgradability and, most of all, reproducibility and methodology. Since TENDL has been seamlessly adopted by many different applications (accelerator, astrophysics, fusion, fission, medical) that require multi-faceted nuclear reaction data forms for shielding, radio-protection, transmutation, materials or earth sciences, it is felt necessary to understand its strengths and remaining weaknesses. The essential knowledge is not the TENDL library itself, but rather the necessary methods, processes, codes, tools and know-how that go into the making of every evaluations of such a library. Current efforts are focused on a proper evaluation of the underlying physics and incorporation of information and metrics into the scientific TALYS/T6 system. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

International Intercomparison Exercise for Nuclear Accident Dosimetry at the DAF Using GODIVA-IV

The Nuclear Criticality Safety Program operated under the direction of Dr. Jerry McKamy completed the first NNSA Nuclear Accident Dosimetry exercise on May 27, 2016. Participants in the exercise were from Lawrence Livermore National Laboratory (LLNL), Los Alamos National Laboratory (LANL), Sandia National Laboratory (SNL), Savanah River Site (SRS), Pacific Northwest National Laboratory (PNNL), US Navy, the Atomic Weapons Establishment (United Kingdom) under the auspices of JOWOG 30, and the Institute for Radiological Protection and Nuclear Safety (France) by special invitation and NCSP memorandum of understanding. This exercise was the culmination of a series of Integral Experiment Requests (IER) that included the establishment of the Nuclear Criticality Experimental Research Center, (NCERC) the startup of the Godiva Reactor (IER-194), the establishment of a the Nuclear Accident Dosimetry Laboratory (NAD LAB) in Mercury, NV, and the determination of reference dosimetry values for the mixed neutron and photon radiation field of Godiva within NCERC.

61 RADIATION PROTECTION AND DOSIMETRY↗

Low-Z linac targets for low-MV gold nanoparticle radiation therapy

Purpose: To investigate the potential of low-Z/low-MV (low-Z) linac targets for gold nanoparticle radiotherapy (GNPT) and to determine the microscopic dose enhancement ratio (DER) due to GNP for the alternative beamlines. In addition, to evaluate the degradation of dose enhancement arising from the increased attenuation of x rays and larger skin dose in water for the low-MV beams compared to the standard linac. Methods: Monte Carlo simulations were used to compute dose and DER for various flattening-filter-free beams (2.5, 4, 6.5 MV). Target materials were beryllium, diamond, and tungsten-copper high-Z target. Target thicknesses were selected based on 20%, 60%, 70%, and 80% of the continuous slowing down approximation electron ranges for a given target material and energy. Evaluation of the microscopic DER was carried out for 100 nm GNP including the degradation factors due to beam attenuation. Results: The greatest increase in DER compared to the standard 6.5 MV linac was for a 2.5 MV Be-target (factor of ∼2). Skin dose ranged from ∼10% (Be, 6.5 MV-80%) to ∼85% (Be, 2.5 MV-20%) depending on the target case. Attenuation of 2.5 MV beams at 22 cm was higher by ∼75% compared with the standard beam. Taking into account the attenuation at 22 cm depth, the effective dose enhancement was up to ∼60% above the DER of the high-Z target. For these cases the effective DER ranged between ∼1.6 and 6 compared with the standard linac. Conclusions: Low-Z (2.5 MV) GNPT is possible even after accounting for greater beam attenuation for deep-seated tumors (22 cm) and the increased skin dose. Further, it can lead to significant sparing of normal tissue while simultaneously escalating the dose in the tumor cells.

61 RADIATION PROTECTION AND DOSIMETRY↗

Progressive cone beam CT dose control in image-guided radiation therapy

Purpose: Cone beam CT (CBCT) in image-guided radiotherapy (IGRT) offers a tremendous advantage for treatment guidance. The associated imaging dose is a clinical concern. One unique feature of CBCT-based IGRT is that the same patient is repeatedly scanned during a treatment course, and the contents of CBCT images at different fractions are similar. The authors propose a progressive dose control (PDC) scheme to utilize this temporal correlation for imaging dose reduction. Methods: A dynamic CBCT scan protocol, as opposed to the static one in the current clinical practice, is proposed to gradually reduce the imaging dose in each treatment fraction. The CBCT image from each fraction is processed by a prior-image based nonlocal means (PINLM) module to enhance its quality. The increasing amount of prior information from previous CBCT images prevents degradation of image quality due to the reduced imaging dose. Two proof-of-principle experiments have been conducted using measured phantom data and Monte Carlo simulated patient data with deformation. Results: In the measured phantom case, utilizing a prior image acquired at 0.4 mAs, PINLM is able to improve the image quality of a CBCT acquired at 0.2 mAs by reducing the noise level from 34.95 to 12.45 HU. In the synthetic patient case, acceptable image quality is maintained at four consecutive fractions with gradually decreasing exposure levels of 0.4, 0.1, 0.07, and 0.05 mAs. When compared with the standard low-dose protocol of 0.4 mAs for each fraction, an overall imaging dose reduction of more than 60% is achieved. Conclusions: PINLM-PDC is able to reduce CBCT imaging dose in IGRT utilizing the temporal correlations among the sequence of CBCT images while maintaining the quality.

61 RADIATION PROTECTION AND DOSIMETRY↗

SU-F-I-06: Evaluation of Imaging Dose for Modulation Layer Based Dual Energy Cone-Beam CT

Purpose: Dual energy cone beam CT system is finding a variety of promising applications in diagnostic CT, both in imaging of endogenous materials and exogenous materials across a range of body sites. Dual energy cone beam CT system to suggest in this study acquire image by rotating 360 degree with half of the X-ray window covered using copper modulation layer. In the region that covered by modulation layer absorb the low energy X-ray by modulation layer. Relative high energy X-ray passes through the layer and contributes to image reconstruction. Dose evaluation should be carried out in order to utilize such an imaging acquirement technology for clinical use. Methods: For evaluating imaging dose of modulation layer based dual energy cone beam CT system, Prototype cone beam CT that configured X-ray tube (D054SB, Toshiba, Japan) and detector (PaxScan 2520V, Varian Medical Systems, Palo Alto, CA) is used. A range of 0.5–2.0 mm thickness of modulation layer is implemented in Monte Carlo simulation (MCNPX, ver. 2.6.0, Los Alamos National Laboratory, USA) with half of X-ray window covered. In-house phantom using in this study that has 3 cylindrical phantoms configured water, Teflon air with PMMA covered for verifying the comparability the various material in human body and is implemented in Monte Carlo simulation. The actual dose with 2.0 mm copper covered half of X-ray window is measured using Gafchromic EBT3 film with 5.0 mm bolus for compared with simulative dose. Results: Dose in phantom reduced 33% by copper modulation layer of 2.0 mm. Scattering dose occurred in modulation layer by Compton scattering effect is 0.04% of overall dose. Conclusion: Modulation layer of that based dual energy cone beam CT has not influence on unnecessary scatter dose. This study was supported by the Radiation Safety Research Programs (1305033) through the Nuclear Safety and Security Commission.

60 APPLIED LIFE SCIENCES↗

Protection factors for French dwellings - 25573

IRSN's current method of assessment of public exposure from atmospheric releases of radioactivity from nuclear installations assumes an exposed person outside of any building; even though the majority of time is spent inside. A more realistic assessment would take into account the protection from atmospheric contamination and deposits provided by dwellings. Therefore, IRSN carried out a study to estimate protection factors for gamma ray emitting radionuclides. This was done with the MCNP calculation code and assumed several French dwelling types. Air contamination (outside and inside) and deposits on soil (outside and inside) and on a roof are considered. Method and results are set out for a few radionuclides. The results show the influence of dwelling type, and also of room sealing (considering inner contamination or not) in case of radiological emergency. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗

Simulation of the Post-Shot Radiation Environment in the National Ignition Facility

The National Ignition Facility at Lawrence Livermore National Laboratory is the world’s biggest and most energetic laser system for inertial confinement fusion. The NIF is designed to perform shots with varying fusion yield (up to 20 MJ or 7.1 × 10 18 neutrons per shot). A great number of diagnostic instruments are present inside the target chamber (TC) and target bay (TB) during shots. The gamma dose rates due to neutron activation are estimated at various decay times following the high-yield (20-MJ) shots. Several components, like the snout assemblies of the diagnostic instrument manipulators and target positioners are inserted inside the TC, close to the target during the shot. These components represent major sources of gamma decay after retraction outside the TC. Five days after a 20-MJ shot, dose rates near the highly activated (retracted) parts are on the order of 1 mSv/h and dose rates within the TB outside the TC but at distance from the retracted components drop to about 50 to 70 μSv/h. The dose is dominated by decay of 24 Na (T 1/2 = 14.95 h) and waiting for two additional days drops the dose rates significantly. Seven days following a 20-MJ shot, dose rates in the immediate vicinity of the retracted components drop to <0.2 mSv/h and the general ambient dose rates within the TB (away from retracted components) near the TC drop to <10 μSv/h. Dose rates at much larger distances from the TC (near TB wall) are an order of magnitude lower. Detailed radiation transport simulations are performed to create detailed dose rate maps for all floors inside the TB. The maps are used to measure worker stay-out times following shots before entry is permitted into the TB.

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SU-E-T-569: Neutron Shielding Calculation Using Analytical and Multi-Monte Carlo Method for Proton Therapy Facility

Purpose: To evaluate the shielding wall design to protect patients, staff and member of the general public for secondary neutron using a simply analytic solution, multi-Monte Carlo code MCNPX, ANISN and FLUKA. Methods: An analytical and multi-Monte Carlo method were calculated for proton facility (Sumitomo Heavy Industry Ltd.) at Samsung Medical Center in Korea. The NCRP-144 analytical evaluation methods, which produced conservative estimates on the dose equivalent values for the shielding, were used for analytical evaluations. Then, the radiation transport was simulated with the multi-Monte Carlo code. The neutron dose at evaluation point is got by the value using the production of the simulation value and the neutron dose coefficient introduced in ICRP-74. Results: The evaluation points of accelerator control room and control room entrance are mainly influenced by the point of the proton beam loss. So the neutron dose equivalent of accelerator control room for evaluation point is 0.651, 1.530, 0.912, 0.943 mSv/yr and the entrance of cyclotron room is 0.465, 0.790, 0.522, 0.453 mSv/yr with calculation by the method of NCRP-144 formalism, ANISN, FLUKA and MCNP, respectively. The most of Result of MCNPX and FLUKA using the complicated geometry showed smaller values than Result of ANISN. Conclusion: The neutron shielding for a proton therapy facility has been evaluated by the analytic model and multi-Monte Carlo methods. We confirmed that the setting of shielding was located in well accessible area to people when the proton facility is operated.

61 RADIATION PROTECTION AND DOSIMETRY↗

Evaluation of dispersion strengthened nickel-base alloy heat shields for space shuttle application

The results obtained in a program to evaluate dispersion-strengthened nickel-base alloys for use in a metallic radiative thermal protection system operating at surface temperatures to 1477 K for the space shuttle were presented. Vehicle environments having critical effects on the thermal protection system are defined; TD Ni-20Cr characteristics of material used in the current study are compared with previous results; cyclic load, temperature, and pressure effects on sheet material residual strength are investigated; the effects of braze reinforcement in improving the efficiency of spotwelded joints are evaluated; parametric studies of metallic radiative thermal protection systems are reported; and the design, instrumentation, and testing of full scale subsize heat shield panels in two configurations are described. Initial tests of full scale subsize panels included simulated meteoroid impact tests, simulated entry flight aerodynamic heating, programmed differential pressure loads and temperatures simulating mission conditions, and acoustic tests simulating sound levels experienced during boost flight.

Johnson, R., Jr.↗

Determination of activation products and resulting decommissioning implications for the Varian TruBeam linear accelerator - Paper 52

Modern medical linear accelerators used for the treatment of cancers are capable of operating at electron energies as high as 20 MeV. Above electron energies of 6 MeV, a measurable dose rate is obtained at the isocenter after beam termination. This is attributed to the activation of linear accelerator components through photonuclear reactions and neutron capture. Photons are produced as Bremsstrahlung during the collision of high-energy electrons with a heavy metal target, typically comprised of copper or tungsten. When photons exceeding the nucleon binding energy interact with a nucleus, neutrons are ejected. In the process of neutron capture, neutrons produced through ({gamma}, n) reactions are absorbed by a nucleus and the excess energy is emitted as a photon. Over the lifetime of an accelerator many components, such as the target, flattening filter, collimators, jaws, and treatment couch, are activated through these interactions. Although dependent on operating conditions, linear accelerators have an expected lifetime of 7 to 10 years. In Canada, at the end of their service, accelerators are typically sold for scrap, returned to a remanufacturing enterprise, or if the machine is still operational it can be retasked for continued use. Since many of the accelerator components have been exposed to a neutron fluence and activated to various extents, the requirements of the Canadian Nuclear Substance and Radiation Device regulations apply prior to final removal from the hospital. For shipments to the USA, there are exemption regulations to be considered. NRC regulation 10CFR 30.71 Schedule B lists a number of common radionuclides and the specific exemption quantity. Below these levels, radioactive materials may be freely transported from Canada to the USA without prior licensing from the Canadian Nuclear Safety Commission (CNSC) or United States Nuclear Regulatory Commission (NRC). (authors)

43 PARTICLE ACCELERATORS↗

SAVY O-ring characterization for lifetime extension

Los Alamos National Laboratory and Nuclear Filter Technology, Inc. developed the SAVY-4000 nuclear material storage container, named after the designers Stone, Anderson, Viers, and Yarbro. The SAVY 4000 meets the requirements of Department of Energy (DOE) M 441.1-1,2 Nuclear Material Packaging Manual, to protect workers who handle nuclear materials against radiation exposures resulting from the leakage of stored materials. The SAVY-4000 is an innovative and creative design: it can be opened and closed in a few seconds without using torque wrenches and instead only a tire iron or equivalent prying tool; has a built-in, fire-rated filter that prevents the build-up of hydrogen gas yet retains 99.97% of plutonium particulates at 200 ACCM rated flow; and prevents the release of stored materials upon a 12- foot drop.

42 ENGINEERING↗

Summary of Building Protection Factor Studies for External Exposure to Ionizing Radiation

Radiation dose assessments are used to help inform decisions to minimize health risks in the event of an atmospheric release of radioactivity including, for example, from a Radiological Dispersal Device, an Improvised Nuclear Device detonation, or a Nuclear Power Plant accident. During these incidents, radiation dose assessments for both indoor and outdoor populations are needed to make informed decisions. These dose assessments inform emergency plans and decisions including, for example, identifying areas in which people should be sheltered and determining when controlled population evacuations should be made. US dose assessment methodologies allow consideration of the protection, and therefore dose reduction, that buildings provide their occupants. However, these methodologies require an understanding of the protection provided by various building types that is currently lacking. To help address this need, Lawrence Livermore National Laboratory, in cooperation with Sandia National Laboratories and the Nuclear Regulatory Commission, was tasked with (a) identifying prior building protection studies, (b) extracting results relevant to US building construction, and (c) summarizing building protection by building type. This report focuses primarily on the protection against radiation from outdoor fallout particles (external gamma radiation).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

SU-E-T-612: Photonuclear Activation of Prosthetic Hips in Radiotherapy

Purpose: To measure the neutron induced activation of a prosthetic hip when exposed to an 18 MV radiotherapy linac beam to assess the potential dose to patients. Methods: A prosthetic hip (Thackray, UK) was placed in a water phantom and irradiated (both in-field and out-of-field) with an 18 MV linac beam. Gamma spectroscopy was used to identify the radioisotopes produced. Following the in-field irradiation where the induced activity is higher, high sensitivity lithium fluoride Thermoluminescence Dosimeters (TLD-100H) (Harshaw, USA) were placed on the surface to measure the dose which would be deposited to nearby tissue resulting from the induced radioactivity. Results: The radioisotopes produced in the hip prosthesis have been identified as {sup 5{sup 2}}V, {sup 5{sup 3}}Fe, and {sup 56}Mn which have half-lives of 3.74, 8.51 and minutes respectively. The 378 and 511 keV characteristic peaks of {sup 5{sup 3}}Fe do not appear in the spectra collected from hips irradiated in the out-of-field region. This isotope is produced by photonuclear interactions within the hip itself. {sup 52}V and {sup 56}Mn appear in spectra collected from hips irradiated either within the photon beam or in the nearby out-of-field region, implying that they are produced by incident neutron radiation produced in components of the linear accelerator.The integrated dose which would be deposited in the tissue immediately surrounding the hip after four hours (1.5 half-lives of the longest lived product) was measured as approximately 100 µGy, following a 10,000 MU irradiation of the prosthesis. A highly accurate measurement of the dose is challenging because of the irregular shape of the prosthesis. Conclusion: The cumulative dose measured at the surface of the hip is 0.0001% of the in-field dose, and is therefore negligible compared with the doses the patient receives as a consequence of their treatment.

61 RADIATION PROTECTION AND DOSIMETRY↗

Nuclear Safety [Vol. 32, No. 1, January-March 1991]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 The Safety Review and Approval Process for Space Nuclear Power Sources, G. L. Bennett; 19 Report on the American Nuclear Society International Topical Meeting: "The Safety, Status, and Future of Non-Commercial Reactors and Irradiation Facilities", E. G. Silver; 35 Book Review: Fission Product Transport Processes in Reactor Accidents, Proceedings to the International Centre tor Heat and Mass Processes T. S. Kress; 38 Fast Reactor Technology in the 1990s: A Summary of the 1990 International Fast Reactor Safety Meeting, A. E. Levin; ACCIDENT ANALYSIS: 56 Effects of Chemical Phenomena on LWR Severe Accident Fission Product Behavior, A. P. Malinauskas and T. S. Kress; CONTROL AND INSTRUMENTATION: 65 Technical Note: Safety Parameter Display Systems—10 Years Later, R. J. Eckenrode; 68 Potential Application of Neural Networks to the Operation of Nuclear Power Plants, R. E. Uhrig; DESIGN FEATURES: 80 Twenty-First DOE/NRC Nuclear Air-Cleaning Conference, R. R. Bellamy, D. W. Moeller, and M. W. First; 91 Impact of an Apparent Radiation Embrittlement Rate on the Life Expectancy of PWR Vessel Supports, R. D. Cheverton, G. C. Robinson, W. E. Pennell, and R. K. Nanstad; ENVIRONMENTAL EFFECTS: 103 Technical Note: The Impact of Offsite Factors on the Safety Performance of Small Nuclear Power Plants, Yu. D. Baranaev and A. N. Viktorov; WASTE AND SPENT FUEL MANAGEMENT: 109 Activities Related to Waste Management, Compiled by E. G. Silver; OPERATING EXPERIENCES: 118 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 121 Selected Safety-Related Events, Compiled by G. A. Murphy; 123 Operating U.S. Power Reactors, Compiled by E. G. Silver; RECENT DEVELOPMENTS: 140 General Administrative Activities, Compiled by E. G. Silver; 150 Reports, Standards, and Safety Guides, D. S. Queener; 155 Status of Power-Reactor Licensing Activities, Compiled by E. G. Silver; 157 Proposed Rule Changes as of Sept. 30, 1990; ANNOUNCEMENTS: 79 MIT Offers Summer Program on Nuclear Power Reactor Safety; 108 Harvard School of Public Health Offers Several Short Courses; 139 Short Course and Workshop on Nuclear Criticality Safety at University of New Mexico; 160 SCK/CEN Announces Training Course on Emergency Planning and Response; 161 The Authors; 164 Indexes to Nuclear Safety, Volume 31

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Small field in-air output factors: The role of miniphantom design and dosimeter type

Purpose: The commissioning of treatment planning systems and beam modeling requires measured input parameters. The measurement of relative output in-air, S{sub c} is particularly difficult for small fields. The purpose of this study was to investigate the influence of miniphantom design and detector selection on measured S{sub c} values for small fields and to validate the measurements against Monte Carlo simulations. Methods: Measurements were performed using brass caps (with sidewalls) or tops (no sidewalls) of varying heights and widths. The performance of two unshielded diodes (60012 and SFD), EBT2 radiochromic film, and a fiber optic dosimeter (FOD) were compared for fields defined by MLCs (5–100 mm) and SRS cones (4–30 mm) on a Varian Novalis linear accelerator. Monte Carlo simulations were performed to theoretically predict S{sub c} as measured by the FOD. Results: For all detectors, S{sub c} agreed to within 1% for fields larger than 10 mm and to within 2.3% for smaller fields. Monte Carlo simulation matched the FOD measurements for all size of cone defined fields to within 0.5%. Conclusions: Miniphantom design is the most important variable for reproducible and accurate measurements of the in-air output ratio, S{sub c}, in small photon fields (less than 30 mm). Sidewalls are not required for fields ≤ 30 mm and tops are therefore preferred over the larger caps. Unlike output measurements in water, S{sub cp,} the selection of detector type for S{sub c} is not critical, provided the active dosimeter volume is small relative to the field size.

61 RADIATION PROTECTION AND DOSIMETRY↗

Integral dose investigation of non-coplanar treatment beam geometries in radiotherapy

Purpose: Automated planning and delivery of non-coplanar plans such as 4π radiotherapy involving a large number of fields have been developed to take advantage of the newly available automated couch and gantry on C-arm gantry linacs. However, there is an increasing concern regarding the potential changes in the integral dose that needs to be investigated. Methods: A digital torso phantom and 22 lung and liver stereotactic body radiation therapy (SBRT) patients were included in the study. The digital phantom was constructed as a water equivalent elliptical cylinder with a major axis length of 35.4 cm and minor axis of 23.6 cm. A 4.5 cm diameter target was positioned at varying depths along the major axis. Integral doses from intensity modulated, non-coplanar beams forming a conical pattern were compared against the equally spaced coplanar beam plans. Integral dose dependence on the phantom geometry and the beam number was also quantified. For the patient plans, the non-coplanar and coplanar beams and fluences were optimized using a column generation and pricing approach and compared against clinical VMAT plans using two full (lung) or partial coplanar arcs (liver) entering at the side proximal to the tumor. Both the average dose to the normal tissue volume and the total volumes receiving greater than 2 Gy (V2) and 5 Gy (V5) were evaluated and compared. Results: The ratio of integral dose from the non-coplanar and coplanar plans depended on the tumor depth for the phantom; for tumors shallower than 10 cm, the non-coplanar integral doses were lower than coplanar integral doses for non-coplanar angles less than 60°. Similar patterns were observed in the patient plans. The smallest non-coplanar integral doses were observed for tumor 6–8 cm deep. For the phantom, the integral dose was independent of the number of beams, consistent with the liver SBRT patients but the lung SBRT patients showed slight increase in the integral dose when more beams were used. Larger tumor size and larger patient body size did not change the overall relationship of integral doses between non-coplanar and coplanar cases. However, the thin disk-shaped tumor received at least 40% greater integral doses with the non-coplanar plans. Overall, patient non-coplanar integral doses and V5 were comparable to those of coplanar doses from the same optimization engine and 15%–20% lower than state of the art VMAT plans. However, non-coplanar beams significantly increased V2 in both the phantom and patients. On average, the lung and liver SBRT patient normal tissue volumes receiving dose greater than 2 Gy were increased by 749 and 532 cm{sup 3}, respectively. Conclusions: The authors used a digital phantom simulating a patient torso and 22 SBRT patients to show that the integral doses from the plans employing optimized non-coplanar beams are comparable to those of the coplanar plans using an equal number of discrete beams and are significantly lower than those of VMAT plans. The non-coplanar beams expose a larger normal tissue volume to non-zero doses, whose impact will need to be evaluated individually to determine the risk/benefit ratio of the non-coplanar plans.

61 RADIATION PROTECTION AND DOSIMETRY↗