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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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At least 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

SU-E-J-69: Iterative Deconvolution of the Initial Photon Fluence for EPID Dosimetry: A Monte Carlo Based Study

Purpose: Developing a fast and accurate calculation model to reconstruct the applied photon fluence from an external photon radiation therapy treatment based on an image recorded by an electronic portal image device (EPID). Methods: To reconstruct the initial photon fluence the 2D EPID image was corrected for scatter from the patient/phantom and EPID to generate the transmitted primary photon fluence. This was done by an iterative deconvolution using precalculated point spread functions (PSF). The transmitted primary photon fluence was then backprojected through the patient/phantom geometry considering linear attenuation to receive the initial photon fluence applied for the treatment.The calculation model was verified using Monte Carlo simulations performed with the EGSnrc code system. EPID images were produced by calculating the dose deposition in the EPID from a 6 MV photon beam irradiating a water phantom with air and bone inhomogeneities and the ICRP anthropomorphic voxel phantom. Results: The initial photon fluence was reconstructed using a single PSF and position dependent PSFs which depend on the radiological thickness of the irradiated object. Appling position dependent point spread functions the mean uncertainty of the reconstructed initial photon fluence could be reduced from 1.13 % to 0.13 %. Conclusion: This study presents a calculation model for fluence reconstruction from EPID images. The{sup Result} show a clear advantage when position dependent PSF are used for the iterative reconstruction. The basic work of a reconstruction method was established and further evaluations must be made in an experimental study.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2015

WE-AB-BRB-07: Alanine and Monte Carlo Determined Beam Quality Corrections for Nonstandard Fields of the Varian TrueBeam Accelerator

Purpose: To determine beam quality correction factors for flattened and unflattened beams of the Varian TrueBeam™ accelerator using alanine measurements and Monte Carlo calculations. Methods: Measurements were performed using L-α-alanine pellets encased in Virtual Water™ paddles. These were irradiated in liquid water using various field sizes of the 6MV and 10MV beam energies of the TrueBeam™, with both flattening-filter-free (FFF) and flattened beam modes. Measurements were also performed in a 10×10 cm{sup 2} {sup 60}Co field. Pellets were read out by the National Physical Laboratory (NPL, Teddington, UK) using electron paramagnetic resonance (EPR) spectrometry, and results were corrected for energy dependence using graphite calorimetry. Beam quality correction factors (k{sub Q}, k{sub Q}msr,Q, and k{sub Q}clin,Q{sub msr}) were determined for three ionization chambers: an Exradin A12 Farmer-type chamber, an Exradin A1SL scanning chamber, and an Exradin A26 microchamber. All chambers were modeled in the egs-chamber user code of EGSnrc. Sources were created using Varian-supplied IAEA-compliant phase spaces and the BEAMnrc user code, and were validated by comparing measured and simulated dose profiles. Results: Measured and calculated k{sub Q} values agreed within uncertainties, showing the chamber models to be reliable. A comparison of measured and calculated k{sub Q}msr,Q results with TG-51-based values showed that TG-51 adequately accounts for variations in beam quality between flattened and unflattened 10×10 cm{sup 2} fields. Small field corrections, k{sub Q}clin,Q{sub msr}, were determined to be up to 5%, referenced to a 10×10 cm{sup 2} field of the same energy and mode. Conclusion: Beam quality corrections were determined for several beam energies of the TrueBeam™ accelerator using Monte Carlo calculations and were validated using measurements with alanine. Values of k{sub Q} determined using TG-51 were found to be adequate for FFF reference fields. However, small field corrections are necessary for both flattened and unflattened beams to ensure greater treatment planning accuracy.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2012

Intrinsic Radiation Source Generation with the ISC Package: Data Comparisons and Benchmarking

The characterization of radioactive emissions from unstable isotopes (intrinsic radiation) is necessary for shielding and radiological-dose calculations from radioactive materials. While most radiation transport codes, e.g., MCNP [X-5 Monte Carlo Team, 2003], provide the capability to input user prescribed source definitions, such as radioactive emissions, they do not provide the capability to calculate the correct radioactive-source definition given the material compositions. Special modifications to MCNP have been developed in the past to allow the user to specify an intrinsic source, but these modification have not been implemented into the primary source base [Estes et al., 1988]. To facilitate the description of the intrinsic radiation source from a material with a specific composition, the Intrinsic Source Constructor library (LIBISC) and MCNP Intrinsic Source Constructor (MISC) utility have been written. The combination of LIBISC and MISC will be herein referred to as the ISC package. LIBISC is a statically linkable C++ library that provides the necessary functionality to construct the intrinsic-radiation source generated by a material. Furthermore, LIBISC provides the ability use different particle-emission databases, radioactive-decay databases, and natural-abundance databases allowing the user flexibility in the specification of the source, if one database is preferred over others. LIBISC also provides functionality for aging materials and producing a thick-target bremsstrahlung photon source approximation from the electron emissions. The MISC utility links to LIBISC and facilitates the description of intrinsic-radiation sources into a format directly usable with the MCNP transport code. Through a series of input keywords and arguments the MISC user can specify the material, age the material if desired, and produce a source description of the radioactive emissions from the material in an MCNP readable format. Further details of using the MISC utility can be obtained from the user guide [Solomon, 2012]. The remainder of this report presents a discussion of the databases available to LIBISC and MISC, a discussion of the models employed by LIBISC, a comparison of the thick-target bremsstrahlung model employed, a benchmark comparison to plutonium and depleted-uranium spheres, and a comparison of the available particle-emission databases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Worst Case Sabotage Consequence Modeling for Advanced Reactors

Under proposed rulemaking by the United States Nuclear Regulatory Commission, advanced reactor licensees may qualify for a reduction in onsite responders if the consequences from a sabotage attack do not exceed 25 rem at the reactor site boundary 2 hours after a release occurs. This work presents a preliminary assessment of the feasibility of meeting the 25 rem at the site boundary criteria. Using advanced reactor radionuclide inventories radiological release estimates created by collaborators at Oak Ridge National Laboratory, the MACCS consequence analysis code was used to investigate the distance at the dose from hypothetical sabotage events reaches 25 rem for different advanced reactor technologies. It was found that the distance to 25 rem depended heavily on sampled weather conditions and sabotage release fractions, and further refinement of release fractions and MACCS modeling parameters is recommended.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Containment margins in FFTF for postulated failure of in-vessel post-accident heat removal

This study was performed to complete the assessment of potential melt-through consequences for the postulated failure of post-accident heat removal following a hypothetical core disruptive accident (HCDA) in the FFTF. The SPRAY, SOFIRE, CACECO, HAA-3B, and COMRADEX codes were used to determine potential site boundary radiation doses. Parametric comparisons were made to evaluate the effects of natural hydrogen-oxygen recombination, reactor cavity liner failures and containment coolers. Containment pressure relief, vent/purge options, and the effects of exhaust filtration were also included in the study. The potential site boundary radiation exposures for all options considered were within 10 CFR 100 guideline values.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

SU-E-T-37: A GPU-Based Pencil Beam Algorithm for Dose Calculations in Proton Radiation Therapy

Purpose: Recent developments in radiation therapy have been focused on applications of charged particles, especially protons. Over the years several dose calculation methods have been proposed in proton therapy. A common characteristic of all these methods is their extensive computational burden. In the current study we present for the first time, to our best knowledge, a GPU-based PBA for proton dose calculations in Matlab. Methods: In the current study we employed an analytical expression for the protons depth dose distribution. The central-axis term is taken from the broad-beam central-axis depth dose in water modified by an inverse square correction while the distribution of the off-axis term was considered Gaussian. The serial code was implemented in MATLAB and was launched on a desktop with a quad core Intel Xeon X5550 at 2.67GHz with 8 GB of RAM. For the parallelization on the GPU, the parallel computing toolbox was employed and the code was launched on a GTX 770 with Kepler architecture. The performance comparison was established on the speedup factors. Results: The performance of the GPU code was evaluated for three different energies: low (50 MeV), medium (100 MeV) and high (150 MeV). Four square fields were selected for each energy, and the dose calculations were performed with both the serial and parallel codes for a homogeneous water phantom with size 300×300×300 mm3. The resolution of the PBs was set to 1.0 mm. The maximum speedup of ∼127 was achieved for the highest energy and the largest field size. Conclusion: A GPU-based PB algorithm for proton dose calculations in Matlab was presented. A maximum speedup of ∼127 was achieved. Future directions of the current work include extension of our method for dose calculation in heterogeneous phantoms.

60 APPLIED LIFE SCIENCES↗
NASA NTRS2021

PolyFit: A C++ code for Polynomial Curve Fit with Calculation of Error Bars

In radiobiology, many dose-response results are modeled using the so-called linear-quadratic(LQ)model, which means that results are modeled as a function of dose Das 𝑅(𝐷)=𝛽0+𝛽1𝐷+𝛽2𝐷2. The coefficients 𝛽0, 𝛽1and 𝛽2are obtained from fitting a series of data points (𝑥𝑖,𝑦𝑖), which is usually done using a least-square method. The LQ and more generally the polynomial fit capability is implemented in many software that analyzes data. However, it is often convenient to do the fitting programmatically, especially when a large number of datasets should be analyzed. Furthermore, depending on the software used, some features may not be implemented. In this mini-review, I discuss the basis of polynomial fitting, including the calculation of errors on the coefficients and results, use of weighting and fixing the intercept value (the coefficient 𝛽0). A simple C++ code to perform the polynomial curve fitting is also provided. This code should be useful not only in radiobiology but in other fields of science as well.

Ianik Plante↗
DOE OSTI2014

SU-E-T-03: 3D GPU-Accelerated Secondary Checks of Radiation Therapy Treatment Plans

Purpose: Redundant treatment verifications in conformal and intensity-modulated radiation therapy techniques are traditionally performed with single point calculations. New solutions can replace these checks with 3D treatment plan verifications. This work describes a software tool (Mobius3D, Mobius Medical Systems) that uses a GPU-accelerated collapsed cone algorithm to perform 3D independent verifications of TPS calculations. Methods: Mobius3D comes with reference beam models for common linear accelerators. The system uses an independently developed collapsed cone algorithm updated with recent enhancements. 144 isotropically-spaced cones are used for each voxel for calculations. These complex calculations can be sped up by using GPUs. Mobius3D calculate dose using DICOM information coming from TPS (CT, RT Struct, RT Plan RT Dose). DVH-metrics and 3D gamma tests can be used to compare both TPS and secondary calculations. 170 patients treated with all common techniques as 3DCFRT (including wedged), static and dynamic IMRT and VMAT have been successfully verified with this solution. Results: Calculation times are between 3–5 minutes for 3DCFRT treatments and 15–20 for most complex dMLC and VMAT plans. For all PTVs mean dose and 90% coverage differences are (1.12±0.97)% and (0.68±1.19)%, respectively. Mean dose discrepancies for all OARs is (0.64±1.00)%. 3D gamma (global, 3%/3 mm) analysis shows a mean passing rate of (97.8 ± 3.0)% for PTVs and (99.0±3.0)% for OARs. 3D gamma pasing rate for all voxels in CT has a mean value of (98.5±1.6)%. Conclusion: Mobius3D is a powerful tool to verify all modalities of radiation therapy treatments. Dose discrepancies calculated by this system are in good agreement with TPS. The use of reference beam data results in time savings and can be used to avoid the propagation of errors in original beam data into our QA system. GPU calculations permit enhanced collapsed cone calculations with reasonable calculation times.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2016

Sci-Sat AM: Radiation Dosimetry and Practical Therapy Solutions - 03: Energy dependence of a clinical probe-format calorimeter and its pertinence to absolute photon and electron beam dosimetry

Purpose: To evaluate the intrinsic and absorbed-dose energy dependence of a small-scale graphite calorimeter probe (GPC) developed for use as a routine clinical dosimeter. The influence of charge deposition on the response of the GPC was also assessed by performing absolute dosimetry in clinical linac-based electron beams. Methods: Intrinsic energy dependence was determined by performing constant-temperature calorimetry dose measurements in a water-equivalent solid phantom, under otherwise reference conditions, in five high-energy photon (63.5 < %dd(10){sub X} < 76.3), and five electron (2.3 cm < R{sub 50} < 8.3 cm) beams. Reference dosimetry was performed for all beams in question using an Exradin A19 ion chamber with a calibration traceable to national standards. The absorbed-dose component of the overall energy dependence was calculated using the EGSnrc egs-chamber user code. Results: A total of 72 measurements were performed with the GPC, resulting in a standard error on the mean absorbed dose of better than 0.3 % for all ten beams. For both the photon and electron beams, no statistically-significant energy dependence was observed experimentally. Peak-to-peak, variations in the relative response of the GPC across all beam qualities of a given radiation type were on the order of 1 %. No effects, either transient or permanent, were attributable to the charge deposited by the electron beams. Conclusions: The GPC’s apparent energy-independence, combined with its well-established linearity and dose rate independence, make it a potentially useful dosimetry system capable measuring photon and electron doses in absolute terms at the clinical level.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2021

Dose Rate Analysis of the WCS Consolidated Interim Storage Facility

This report describes a confirmatory calculation for a proposed consolidated interim storage facility (CISF) to support the US Nuclear Regulatory Commission’s review of the license application submitted by the Interim Storage Partners. The scope of this report is limited to an assessment of the annual dose from direct radiation associated with Phase 1 of the Waste Control Specialists (WCS) CISF and the determination of the minimum controlled area boundary based on the 10 CFR 72.104 limit of 25 mrem annual dose to the whole body. The confirmatory dose rate calculations used the source term and shielding calculation capabilities of the SCALE 6.2.3 computer code system. The calculation method employed in the confirmatory calculations uses a detailed Monte Carlo radiation transport simulation from source to dose rate. This method differs from the two-step method used in the safety analysis report (SAR), which requires determination of the photon and neutron energy and angular distributions on the cask external surface and then the use of this surface source in a new radiation transport calculation to determine dose rate as a function of distance from the storage facility. A simulation was made using the complete site geometry (all casks present) but with only one cask containing source, and a total of 467 independent calculations were performed to obtain the dose rate maps produced by each storage cask. Basic input data and assumptions used in the SAR (e.g., cask design parameters and design basis assembly characteristics) were used in this confirmatory calculation for consistency with the SAR. Local atmospheric conditions for skyshine calculations were simulated using US standard atmosphere data, which describe average air temperature, pressure, and density as a function of altitude. The determined minimum required distances from the WCS Phase 1 CISF to a contour defined by the 25 mrem annual dose limit were approximately 623 m (0.387 miles) in the N-NE direction, 594 m (0.369 miles) in the SSW direction, and 533 m (0.331 miles) in the E-SE direction and the W-NW direction. The annual dose results of the confirmatory calculation are in relatively good agreement with the annual dose values presented in the WCS SAR. The annual dose produced by the Phase 1 CISF at the nearest site boundary, 0.75 miles away from the CISF, was estimated by this study to be approximately 6.02E-02 ± 2.0E-04 mrem, which is significantly smaller than the 25 mrem annual dose limit provided in 10 CFR 72.104. The SAR annual dose rate value for the nearest site boundary was 7.52E-2 mrem, which included 7.73E-03 mrem due to postulated leakage of the FO-, FC-, and FF-canisters. A comparison with the dose rate values reported in the SAR for 16 detector locations, D1 through D16, on the protected area boundary is also included in this report. This comparison shows that the SAR determined higher annual dose values for detectors D1 through D14 by 12% to 45% and lower annual dose values for detectors D15 and D16 by 18% and 8%, respectively, as compared with this ORNL confirmatory calculation. Sensitivity studies have been performed to support a comprehensive review of the application. In these studies, the effects on dose rate of the modeled air volume, density, and humidity were analyzed. The sensitivity calculations showed that the air volume used in the Monte Carlo simulations was adequate and that a larger air volume would not further increase the dose rate due to skyshine contributions. The sensitivity analysis showed that simulation of local atmospheric conditions is important for the accuracy of dose rate estimates at large distances from the facility. Air humidity was demonstrated to have negligible effects on dose rate values for the simulated local atmospheric conditions.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2019

Dose Modeling for Harborview Research and Training Facility

The Harborview Research and Training (R&T) Facility at the University of Washington (UW) was accidently contaminated from a breached sealed source of Cesium-137. The release occurred during removal of the sealed source from its holder within a portable hot cell in the loading dock area near room 220. The loading dock and northeast corner of the building above the ground floor were impacted by the release, whereas the remaining portions of the R&T Facility were minimally affected. Detailed descriptions of the R&T Facility and the affected area/rooms are included in Section 2. The RESRAD Team in the Environmental Science Division of Argonne National Laboratory visited the site on August 12, 2019, and interviewed the R&T Facility building manager and staff working in the building at that time. Site- and building-specific data were collected including building floor plan CAD drawings, building occupants’ working hours, contaminated rooms photos and their dimensions, building air exchange rates, building materials and “curtain wall” details, and so forth. These data are being used in the dose modeling for potential occupants of the building after it is remediated. The dose modeling was performed using the RESRAD-BUILD computer code, which is designed for the evaluation of radiologically contaminated buildings (Yu et al. 2003). Two hypothetical building occupants of the Harborview R&T Facility were evaluated: a researcher occupying room 726 (a research laboratory on the seventh floor) and a custodian occupying room 220 (loading dock and trash compactor room with a rollup door). A maintenance worker that will be spending less time in impacted areas was not modeled because it will be bounded by other two hypothetical building occupants modeled. A summary description of the RESRAD-BUILD is provided in Appendix A, and the modeling of potential doses to a researcher and a custodian is described in detail in Section 3. A summary and a discussion of the dose modeling results are included in Section 4. The parameters used in the RESRAD-BUILD modeling are included in Appendix B. This report documents the dose modeling results; it does not describe measurements of the actual radioactivity levels at the Harborview R&T Facility, nor does it set any expectation for the actual final radioactivity levels. The radiation doses presented are modeling results based on the generic screening values of Cs-137 and several conservative assumptions and are meant only to assist in decision making regarding decontamination of the Harborview R&T Facility.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2020

A thermodynamics-based damage model for the non-linear mechanical behavior of SiC/SiC ceramic matrix composites in irradiation and thermal environments

A damage model is developed and validated with experimental data for the non-linear mechanical behavior of SiC/SiC composite materials in nuclear applications. Cyclic thermal and mechanical loading associated with neutron irradiation effects of these composites leads to wide-spread and progressive micro-cracking that leads to loss of thermal conductivity and further enhancement of thermo-mechanical damage. A physics-based model of wide-spread micro-cracking is developed within the thermodynamic framework of continuum damage mechanics. Evolution equations for damage parameters that describe the growth of continuum damage are developed, where the material variables are obtained from experiments. The model novelty is in coupling mechanical, thermal, and irradiation damage through a consistent thermodynamic framework, including loss of thermal conductivity due to the evolution of mechanically induced micro-cracks. A number of thermo-mechanical experiments were conducted to confirm model assumptions. The model is shown to be validated with out-of-pile experiments, and then implemented using commercial finite element code COMSOL to the fuel cladding problem with normal and high radiation dose cases.

Materials Science↗
DOE OSTI2024

Evaluation of Dose Coefficients Implemented in MACCS

A variety of dose conversion factor files (DCF files) have been supplied for use with the MACCS code since it was initially released. For MACCS 4.2, the MACCS DCF files have been updated to include coefficients used in the computation of acute skin doses from within the MACCS software to increase functionality and to add a pseudo-organ to represent the total effective dose equivalent (TEDE) (as defined in 10 CFR 20.1003) based on International Commission on Radiological Protection (ICRP) Publication 30. This report provides a description of how these changes have been implemented and a summary of the various DCF files supplied with MACCS 4.2. The report also provides supplemental discussions to assist the reader in understanding the technical basis for the MACCS DCFs. These supplemental discussions include a summary of basic dosimetry modeling concepts and a brief review of the Federal Guidance Reports (FGRs) upon which MACCS dose coefficients have historically been based.

97 MATHEMATICS AND COMPUTING↗
NASA NTRS1992

Predictions of dose from electrons in space

The objective of the project is to develop a general-purpose, user-friendly computerized database and code package, for the PC as well as larger computers, which can be used for the routine prediction of the absorbed dose from incident electrons and their secondary bremsstrahlung (and from incident protons) as functions of the thickness of aluminum shielding in space. The assumption of homogeneous aluminum shields and of isotropic incident fluxes (at least in a time-averaged sense) allows for the rather reliable conversion of doses in slabs to those in other simple bodies, such as spherical and cylindrical solids and shells. On such a basis, depth-dose data for monoenergetic incident radiation can be generated once-and-for-all from accurate transport calculations, and this database can then be used repeatedly in rapid dose predictions for arbitrary radiation spectra and for a variety of spacecraft sizes and shapes, without recourse to the very time-consuming Monte Carlo calculations. This project entails a thorough updating, extension, and refinement of our earlier SHIELDOSE package, with the goal of a more reliable, fool-proof, and general system.

Seltzer, Stephen M.↗
DOE OSTI2015

SU-E-T-171: Characterization of the New Xoft Axxent Electronic Brachytherapy Source Using PRESAGE Dosimeters

Purpose: To characterize the Xoft Axxent electronic brachytherapy source using PRESAGE™ dosimeters to obtain independent confirmation of TG-43U1 dosimetry values from previous studies and ascertain its reproducibility in HDR brachytherapy. Methods: PRESAGE™ dosimeters are solid, polyurethane-based dosimeters doped with radiochromic leucodyes that produce a linear optical-density response when exposed to radiation. Eight 1-kg dosimeters were scanned prior to irradiation on an optical-CT scanner to eliminate background signal and any optical imperfections from each dosimeter. To quantify potential imaging artifacts due to oversaturated responses in the immediate range of the source, half of the eight dosimeters were cast with a smaller channel diameter of 5.4 mm, and the other half were cast with a larger channel diameter of 15mm. During irradiation, the catheters were placed in the center of each channel. Catheters fit the 5.4mm diameters channels whereas polyurethane plugs were inserted into the larger channels to create a sturdy, immobile catheter which allowed uniform dose distributions. Two dosimeters of each 5.4mm and 15mm were irradiated at either 1517.3 cGy or 2017.5 cGy. Post-irradiation scans were performed within 48 hours of irradiation. A 3D reconstruction based on subtraction of these two images and the relative dose measurements were made using in-house software. Results: Comparing measured radial dose rates with previous results revealed smaller percent errors when PRESAGE™ irradiations were at lower maximum dose. The dosimeters showed small deviations in radial dose function, g{sub p} (r), from previous studies. Among the dosimeters irradiated at 1517.3 cGy, the g{sub p}(r) compared to previous studies fluctuated from 0.0043 to 0.3922. This suggests small fluctuations can drastically change radial dose calculations. Conclusion: The subtraction of pre-irradiation and post-irradiation scans of PRESAGE™ dosimeters using an optical-CT scanner shows promising results in determining 3D dosimetry for Xoft Axxent devices; however, further research is recommended. NIH Grant#: 5-U24-CA081647-13; ROI Grant#: 5R01CA100835.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2015

SU-E-T-139: Automated Daily EPID Exit Dose Analysis Uncovers Treatment Variations

Purpose: To evaluate a fully automated EPID exit dose system for its ability to detect daily treatment deviations including patient setup, delivery, and anatomy changes. Methods: PerFRACTION (Sun Nuclear Corporation) software is a system that uses integrated EPID images taken during patient treatment and automatically pulled from the Aria database and analyzed based on user-defined comparisons. This was used to monitor 20 plans consisting of a total of 859 fields for 18 patients, for a total of 251 fractions. Nine VMAT, 5 IMRT, and 6 3D plans were monitored. The Gamma analysis was performed for each field within a plan, comparing the first fraction against each of the other fractions in each treatment course. A 2% dose difference, 1 mm distance-to-agreement, and 10% dose threshold was used. These tight tolerances were chosen to achieve a high sensitivity to treatment variations. The field passed if 93% of the pixels had a Gamma of 1 or less. Results: Twenty-nine percent of the fields failed. The average plan passing rate was 92.5%.The average 3D plan passing rate was less than for VMAT or IMRT, 84%, vs. an average of 96.2%. When fields failed, an investigation revealed changes in patient anatomy or setup variations, often also leading to variations of transmission through immobilization devices. Conclusion: PerFRACTION is a fully automated system for determining daily changes in dose transmission through the patient that requires no effort other than for the imager panel to be deployed during treatment. A surprising number of fields failed the analysis and can be attributed to important treatment variations that would otherwise not be appreciated. Further study of inter-fraction treatment variations is possible and warranted. Sun Nuclear Corporation provided a license to the software described.

60 APPLIED LIFE SCIENCES↗
NASA NTRS1994

Experiment K-7-41: Radiation Experiments on Cosmos 2044

The Cosmos 2044 biosatellite mission offered the opportunity for radiation measurements under conditions which are seldom available (an inclination of 82.3 degrees and altitude of 294 x 216 km). Measurements were made on the outside of the spacecraft under near-zero shielding conditions. Also, this mission was the first in which active temperature recorders (the ATR-4) were flown to record the temperature profiles of detector stacks. Measurements made on this mission provide a comparison and test for modeling of depth doses and LET spectra for orbital parameters previously unavailable. Tissue absorbed doses from 3480 rad (252 rad/d) down to 0.115 rad (8.33 mrad/d) were measured at different depths (0.0146 and 3.20 g/sq. cm, respectively) with averaged TLD readings. The LET spectra yielded maximum and minimum values of integral flux of 27.3 x 10-4 and 3.05 x 10(exp -4) cm(exp -2).s(exp -1).sr(exp -4) of dose rate of 7.01 and 1.20 mrad/d, and of dose equivalent rate of 53.8 and 11.6 mrem/d, for LET(infinity).H2O is greater than or equal to 4 keV/micro-m. Neutron measurements yielded 0.018 mremld in the thermal region, 0.25 mrem/d in the resonance region and 3.3 mrem/d in the high energy region. The TLD depth dose and LET spectra have been compared with calculations from the modeling codes. The agreement is good but some further refinements are in order. In comparing measurements on Cosmos 2044 with those from previous Cosmos missions (orbital inclinations of 62.8 degrees) there is a greater spread (maximum to minimum) in depth doses and an increased contribution from GCR's, and higher LET particles, in the heavy particle fluxes.

Benton, E. V.↗
DOE OSTI2018

Dose Maps In/Around Target and Test Cell for DONES During the Operation

DONES (DEMO-Oriented Neutron Source) is to be constructed in order to investigate materials that are intended to be used for DEMO (DEMOnstration Power Plant) since they will be irradiated with very high fluxes of neutrons. A deuteron beam will bombard a lithium target situated in the so called Test Cell (TC) shielded with concrete. Monte Carlo simulations by means of the MCNP (Monte Carlo N-Particle) code were carried out to calculate both horizontal and vertical radiation dose maps in and around of the TC facility. Three geometrical configurations of the shielding and two kinds of concrete (ordinary and magnetite ones) were examined. For each case a spatial dose distribution and an accompanying errors map were provided. Advanced techniques of variance reduction were applied to compute doses in and outside of thick concrete walls. On the grounds of the obtained results a final configuration of the shielding was selected.

07 ISOTOPE AND RADIATION SOURCES↗