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All sourcesNASA NTRSDOE OSTI
DOE OSTI2014

On the suitability of ultrathin detectors for absorbed dose assessment in the presence of high-density heterogeneities

Purpose: The aim of this study was to evaluate the suitability of several detectors for the determination of absorbed dose in bone. Methods: Three types of ultrathin LiF-based thermoluminescent dosimeters (TLDs)—two LiF:Mg,Cu,P-based (MCP-Ns and TLD-2000F) and a{sup 7}Li-enriched LiF:Mg,Ti-based (MTS-7s)—as well as EBT2 Gafchromic films were used to measure percentage depth-dose distributions (PDDs) in a water-equivalent phantom with a bone-equivalent heterogeneity for 6 and 18 MV and a set of field sizes ranging from 5×5 cm{sup 2} to 20×20 cm{sup 2}. MCP-Ns, TLD-2000F, MTS-7s, and EBT2 have active layers of 50, 20, 50, and 30 μm, respectively. Monte Carlo (MC) dose calculations (PENELOPE code) were used as the reference and helped to understand the experimental results and to evaluate the potential perturbation of the fluence in bone caused by the presence of the detectors. The energy dependence and linearity of the TLDs’ response was evaluated. Results: TLDs exhibited flat energy responses (within 2.5%) and linearity with dose (within 1.1%) within the range of interest for the selected beams. The results revealed that all considered detectors perturb the electron fluence with respect to the energy inside the bone-equivalent material. MCP-Ns and MTS-7s underestimated the absorbed dose in bone by 4%–5%. EBT2 exhibited comparable accuracy to MTS-7s and MCP-Ns. TLD-2000F was able to determine the dose within 2% accuracy. No dependence on the beam energy or field size was observed. The MC calculations showed that a50 μm thick detector can provide reliable dose estimations in bone regardless of whether it is made of LiF, water or EBT’s active layer material. Conclusions: TLD-2000F was found to be suitable for providing reliable absorbed dose measurements in the presence of bone for high-energy x-ray beams.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2022

Polyethylene density uncertainty in the simulation of neutron detectors

Simulations with the Monte Carlo N-Particle® (MCNP®) code are frequently used to study neutron coincidence counters for international safeguards. Reducing sources of uncertainty enables more accurate simulations, which reduces the need for expensive or impossible measurements. Accurately specifying the density of polyethylene is crucial for accurate simulations. The density of polyethylene ranges from 0.88 to 0.97 g/cm 3 and high-density polyethylene (HDPE) ranges from about 0.944 to 0.965 g/cm 3 . The density of a specific detector should be measured exactly for the best simulations. Here, the density of the blue/white Active Well Coincidence Counter (AWCC) bulk polyethylene was measured to be 0.9594 +- 0.0013 g/cm 3 at 1-sigma confidence. The density was calculated from repeated measurements of the various polyethylene dimensions for volume, and weighing the polyethylene for the mass. The AWCC neutron detection efficiency was simulated for each polyethylene density. The relative range of efficiencies (3σ) was 3.4% for unspecified polyethylene, 0.9% for HDPE, and 0.3% for the measured density. The impact of modeling small features such as screws was also studied and was found to be a negligible 0.2%. Measuring polyethylene density can reduce its role in uncertainty from one of the largest to one of the smallest contributors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Monte Carlo simulation of TrueBeam flattening-filter-free beams using Varian phase-space files: Comparison with experimental data

Purpose: Phase-space files for Monte Carlo simulation of the Varian TrueBeam beams have been made available by Varian. The aim of this study is to evaluate the accuracy of the distributed phase-space files for flattening filter free (FFF) beams, against experimental measurements from ten TrueBeam Linacs. Methods: The phase-space files have been used as input in PRIMO, a recently released Monte Carlo program based on thePENELOPE code. Simulations of 6 and 10 MV FFF were computed in a virtual water phantom for field sizes 3 × 3, 6 × 6, and 10 × 10 cm{sup 2} using 1 × 1 × 1 mm{sup 3} voxels and for 20 × 20 and 40 × 40 cm{sup 2} with 2 × 2 × 2 mm{sup 3} voxels. The particles contained in the initial phase-space files were transported downstream to a plane just above the phantom surface, where a subsequent phase-space file was tallied. Particles were transported downstream this second phase-space file to the water phantom. Experimental data consisted of depth doses and profiles at five different depths acquired at SSD = 100 cm (seven datasets) and SSD = 90 cm (three datasets). Simulations and experimental data were compared in terms of dose difference. Gamma analysis was also performed using 1%, 1 mm and 2%, 2 mm criteria of dose-difference and distance-to-agreement, respectively. Additionally, the parameters characterizing the dose profiles of unflattened beams were evaluated for both measurements and simulations. Results: Analysis of depth dose curves showed that dose differences increased with increasing field size and depth; this effect might be partly motivated due to an underestimation of the primary beam energy used to compute the phase-space files. Average dose differences reached 1% for the largest field size. Lateral profiles presented dose differences well within 1% for fields up to 20 × 20 cm{sup 2}, while the discrepancy increased toward 2% in the 40 × 40 cm{sup 2} cases. Gamma analysis resulted in an agreement of 100% when a 2%, 2 mm criterion was used, with the only exception of the 40 × 40 cm{sup 2} field (∼95% agreement). With the more stringent criteria of 1%, 1 mm, the agreement reduced to almost 95% for field sizes up to 10 × 10 cm{sup 2}, worse for larger fields. Unflatness and slope FFF-specific parameters are in line with the possible energy underestimation of the simulated results relative to experimental data. Conclusions: The agreement between Monte Carlo simulations and experimental data proved that the evaluated Varian phase-space files for FFF beams from TrueBeam can be used as radiation sources for accurate Monte Carlo dose estimation, especially for field sizes up to 10 × 10 cm{sup 2}, that is the range of field sizes mostly used in combination to the FFF, high dose rate beams.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

Dosimetric evaluation of new approaches in GRID therapy using nonconventional radiation sources

Purpose: Spatial fractionation of the dose has proven to be a promising approach to increase the tolerance of healthy tissue, which is the main limitation of radiotherapy. A good example of that is GRID therapy, which has been successfully used in the management of large tumors with low toxicity. The aim of this work is to explore new avenues using nonconventional sources: GRID therapy by using kilovoltage (synchrotron) x-rays, the use of very high-energy electrons, and proton GRID therapy. They share in common the use of the smallest possible grid sizes in order to exploit the dose–volume effects. Methods: Monte Carlo simulations (PENELOPE/PENEASY and GEANT4/GATE codes) were used as a method to study dose distributions resulting from irradiations in different configurations of the three proposed techniques. As figure of merit, percentage (peak and valley) depth dose curves, penumbras, and central peak-to-valley dose ratios (PVDR) were evaluated. As shown in previous biological experiments, high PVDR values are requested for healthy tissue sparing. A superior tumor control may benefit from a lower PVDR. Results: High PVDR values were obtained in the healthy tissue for the three cases studied. When low energy photons are used, the treatment of deep-seated tumors can still be performed with submillimetric grid sizes. Superior PVDR values were reached with the other two approaches in the first centimeters along the beam path. The use of protons has the advantage of delivering a uniform dose distribution in the tumor, while healthy tissue benefits from the spatial fractionation of the dose. In the three evaluated techniques, there is a net reduction in penumbra with respect to radiosurgery. Conclusions: The high PVDR values in the healthy tissue and the use of small grid sizes in the three presented approaches might constitute a promising alternative to treat tumors with such spatially fractionated radiotherapy techniques. The dosimetric results presented here support the interest of performing radiobiology experiments in order to evaluate these new avenues.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2014

Absolute x-ray dosimetry on a synchrotron medical beam line with a graphite calorimeter

Purpose: The absolute dose rate of the Imaging and Medical Beamline (IMBL) on the Australian Synchrotron was measured with a graphite calorimeter. The calorimetry results were compared to measurements from the existing free-air chamber, to provide a robust determination of the absolute dose in the synchrotron beam and provide confidence in the first implementation of a graphite calorimeter on a synchrotron medical beam line. Methods: The graphite calorimeter has a core which rises in temperature when irradiated by the beam. A collimated x-ray beam from the synchrotron with well-defined edges was used to partially irradiate the core. Two filtration sets were used, one corresponding to an average beam energy of about 80 keV, with dose rate about 50 Gy/s, and the second filtration set corresponding to average beam energy of 90 keV, with dose rate about 20 Gy/s. The temperature rise from this beam was measured by a calibrated thermistor embedded in the core which was then converted to absorbed dose to graphite by multiplying the rise in temperature by the specific heat capacity for graphite and the ratio of cross-sectional areas of the core and beam. Conversion of the measured absorbed dose to graphite to absorbed dose to water was achieved using Monte Carlo calculations with the EGSnrc code. The air kerma measurements from the free-air chamber were converted to absorbed dose to water using the AAPM TG-61 protocol. Results: Absolute measurements of the IMBL dose rate were made using the graphite calorimeter and compared to measurements with the free-air chamber. The measurements were at three different depths in graphite and two different filtrations. The calorimetry measurements at depths in graphite show agreement within 1% with free-air chamber measurements, when converted to absorbed dose to water. The calorimetry at the surface and free-air chamber results show agreement of order 3% when converted to absorbed dose to water. The combined standard uncertainty is 3.9%. Conclusions: The good agreement of the graphite calorimeter and free-air chamber results indicates that both devices are performing as expected. Further investigations at higher dose rates than 50 Gy/s are planned. At higher dose rates, recombination effects for the free-air chamber are much higher and expected to lead to much larger uncertainties. Since the graphite calorimeter does not have problems associated with dose rate, it is an appropriate primary standard detector for the synchrotron IMBL x rays and is the more accurate dosimeter for the higher dose rates expected in radiotherapy applications.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2018

Distinguishing Research Reactor Fuels Using Neutron Multiplicity - 25615

Neutron measurements are one of the reliable means of interrogation of fissile material. Neutron measurements are used for total counting (pure metal fissile samples) or counting coincidence/multiplicity (for impure fissile samples). Various flavors of neutron measurement are used routinely by the IAEA to determine the fissile mass of an unknown sample. The objective of this work is to design an experiment to distinguish the two different fuel types using neutron multiplicity method. The design of experiment is performed using Monte Carlo N- Particle (MCNP) transport code. General Electric (GE) designed 1-MWt pool-type research reactor at UMass Lowell (UMLRR) uses Materials Testing Reactor (MTR) plate-type fuel consisting of 19 full and 2 partial fuel assemblies, controlled by four boron control blades and one regulating control rod. Two different fuel types are available at the UMLRR facility. In 2008, UMass Lowell obtained slightly burnt fuel assemblies from the decommissioned WPI research reactor. The simulation results discussed in this paper show that within the limits of statistical error, the single and double moments for the WPI and UMLRR fuels can be measured. The two fuels have very similar composition and enrichments resulting in very similar neutron multiplicity moments. This implies that neutron multiplicity measurements alone would not be sufficient to distinguish between the two fuel types at low burnups. The change in the observed single-double ratio between each fuel type is less than 6 % which falls below the uncertainty in the efficiency of the JCC71 slab detectors. The neutron signature from low burnup and low enriched uranium fuel are not strong, however, these can be measured under low background and long measurement times. This would be an important consideration in designing the experiments and parallel gamma measurements will also be planned. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Magnetic Resonance Imaging-Based Radiation-Absorbed Dose Estimation of {sup 166}Ho Microspheres in Liver Radioembolization

Purpose: To investigate the potential of magnetic resonance imaging (MRI) for accurate assessment of the three-dimensional {sup 166}Ho activity distribution to estimate radiation-absorbed dose distributions in {sup 166}Ho-loaded poly (L-lactic acid) microsphere ({sup 166}Ho-PLLA-MS) liver radioembolization. Methods and Materials: MRI, computed tomography (CT), and single photon emission CT (SPECT) experiments were conducted on an anthropomorphic gel phantom with tumor-simulating gel samples and on an excised human tumor-bearing liver, both containing known amounts of {sup 166}Ho-PLLA-MS. Three-dimensional radiation-absorbed dose distributions were estimated at the voxel level by convolving the {sup 166}Ho activity distribution, derived from quantitative MRI data, with a {sup 166}Ho dose point-kernel generated by MCNP (Monte Carlo N-Particle transport code) and from Medical Internal Radiation Dose Pamphlet 17. MRI-based radiation-absorbed dose distributions were qualitatively compared with CT and autoradiography images and quantitatively compared with SPECT-based dose distributions. Both MRI- and SPECT-based activity estimations were validated against dose calibrator measurements. Results: Evaluation on an anthropomorphic phantom showed that MRI enables accurate assessment of local {sup 166}Ho-PLLA-MS mass and activity distributions, as supported by a regression coefficient of 1.05 and a correlation coefficient of 0.99, relating local MRI-based mass and activity calculations to reference values obtained with a dose calibrator. Estimated MRI-based radiation-absorbed dose distributions of {sup 166}Ho-PLLA-MS in an ex vivo human liver visually showed high correspondence to SPECT-based radiation-absorbed dose distributions. Quantitative analysis revealed that the differences in local and total amounts of {sup 166}Ho-PLLA-MS estimated by MRI, SPECT, and the dose calibrator were within 10%. Excellent agreement was observed between MRI- and SPECT-based dose-volume histograms. Conclusions: Quantitative MRI was demonstrated to provide accurate three-dimensional {sup 166}Ho-PLLA-MS activity distributions, enabling localized intrahepatic radiation-absorbed dose estimation by convolution with a {sup 166}Ho dose point-kernel for liver radioembolization treatment optimization and evaluation.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2012

Dependences of mucosal dose on photon beams in head-and-neck intensity-modulated radiation therapy: a Monte Carlo study

Dependences of mucosal dose in the oral or nasal cavity on the beam energy, beam angle, multibeam configuration, and mucosal thickness were studied for small photon fields using Monte Carlo simulations (EGSnrc-based code), which were validated by measurements. Cylindrical mucosa phantoms (mucosal thickness = 1, 2, and 3 mm) with and without the bone and air inhomogeneities were irradiated by the 6- and 18-MV photon beams (field size = 1 Multiplication-Sign 1 cm{sup 2}) with gantry angles equal to 0 Degree-Sign , 90 Degree-Sign , and 180 Degree-Sign , and multibeam configurations using 2, 4, and 8 photon beams in different orientations around the phantom. Doses along the central beam axis in the mucosal tissue were calculated. The mucosal surface doses were found to decrease slightly (1% for the 6-MV photon beam and 3% for the 18-MV beam) with an increase of mucosal thickness from 1-3 mm, when the beam angle is 0 Degree-Sign . The variation of mucosal surface dose with its thickness became insignificant when the beam angle was changed to 180 Degree-Sign , but the dose at the bone-mucosa interface was found to increase (28% for the 6-MV photon beam and 20% for the 18-MV beam) with the mucosal thickness. For different multibeam configurations, the dependence of mucosal dose on its thickness became insignificant when the number of photon beams around the mucosal tissue was increased. The mucosal dose with bone was varied with the beam energy, beam angle, multibeam configuration and mucosal thickness for a small segmental photon field. These dosimetric variations are important to consider improving the treatment strategy, so the mucosal complications in head-and-neck intensity-modulated radiation therapy can be minimized.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

SU-E-T-608: Perturbation Corrections for Alanine Dosimeters in Different Phantom Materials in High-Energy Photon Beams

Purpose: Alanine dosimeters are often used for in-vivo dosimetry purposes in radiation therapy. In a Monte Carlo study the influence of 20 different surrounding/phantom materials for alanine dosimeters was investigated. The investigations were performed in high-energy photon beams, covering the whole range from {sup 60}Co up to 25 MV-X. The aim of the study is the introduction of a perturbation correction k{sub env} for alanine dosimeters accounting for the environmental material. Methods: The influence of different surrounding materials on the response of alanine dosimeters was investigated with Monte Carlo simulations using the EGSnrc code. The photon source was adapted with BEAMnrc to a {sup 60}Co unit and an Elekta (E{sub nom}=6, 10, 25 MV-X) linear accelerator. Different tissue-equivalent materials ranging from cortical bone to lung were investigated. In addition to available phantom materials, some material compositions were taken and scaled to different electron densities. The depth of the alanine detectors within the different phantom materials corresponds to 5 cm depth in water, i.e. the depth is scaled according to the electron density (n{sub e}/n{sub e,w}) of the corresponding phantom material. The dose was scored within the detector volume once for an alanine/paraffin mixture and once for a liquid water voxel. The relative response, the ratio of the absorbed dose to alanine to the absorbed dose to water, was calculated and compared to the corresponding ratio under reference conditions. Results: For each beam quality the relative response r and the correction factor for the environment kenv was calculated. k{sub env}=0.9991+0.0049 *((n{sub e}/n{sub e,w})−0.7659){sup 3} Conclusion: A perturbation correction factor k{sub env} accounting for the phantom environment has been introduced. The response of the alanine dosimeter can be considered independent of the surrounding material for relative electron densities (n{sub e}/n{sub e,w}) between 1 and 1.4. For denser materials such as bone or much less dense surroundings such as lung, a small correction would be appropriate.

62 RADIOLOGY AND NUCLEAR MEDICINE↗
DOE OSTI2014

Design of a backscatter 14-MeV neutron time-of-flight spectrometer for experiments at ITER

Neutron energy spectrometry diagnostics play an important role in present-day experiments related to fusion energy research. Measurements and thorough analysis of the neutron emission from the fusion plasma give information on a number of basic fusion performance quantities, on the condition of the neutron source and plasma behavior. Here we discuss the backscatter Time-of-Flight (bTOF) spectrometer concept as a possible instrument for performing high resolution measurements of 14 MeV neutrons. The instrument is based on two sets of scintillators, a first scatterer exposed to a collimated neutron beam and a second detector set placed in the backward direction. The scintillators of the first set are enriched in deuterium to achieve neutron backscattering. The energy resolution and efficiency of a bTOF instrument have been determined for various geometrical configurations. A preliminary design of optimal geometry for the two scintillator sets has been obtained by Monte Carlo simulations based on the MCNPX code.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
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 OSTI2016

Radiation dose and cancer risk to out-of-field and partially in-field organs from radiotherapy for symptomatic vertebral hemangiomas

Purpose: Vertebral hemangiomas (VHs) are the most common benign tumors of the spine that may cause bone resorption. Megavoltage irradiation is usually the treatment of choice for the management of symptomatic VHs. The current study was conducted to estimate the risk for carcinogenesis from radiotherapy of this benign disease on the basis of the calculated radiation doses to healthy organs. Methods: The Monte Carlo N-particle transport code was employed to simulate the irradiation with 6 MV x-rays of a VH presented in the cervical, upper thoracic, lower thoracic, and lumbar spine. The average radiation dose (D{sub av}) received by each critical organ located outside the primarily irradiated area was calculated. Three-dimensional treatment plans were also generated for the VHs occurring at the four different sites of the spinal cord based on patients’ computed tomography data. The organ equivalent dose (OED) to each radiosensitive structure, which was partly encompassed by the applied treatment fields, was calculated with the aid of differential dose–volume histograms. The D{sub av} and the OED values were combined with a linear-no-threshold model and a nonlinear mechanistic model, respectively, to estimate the organ-, age-, and gender-specific lifetime attributable risks (LARs) for cancer development. The estimated risks were compared with the respective nominal lifetime intrinsic risks (LIRs) for the unexposed population. Results: For a standard target dose of 34 Gy, the OED varied from 0.39–5.15 Gy by the organ of interest and the irradiation site. The D{sub av} range for the out-of-field organs was 4.9 × 10{sup −4} to 0.56 Gy. The LAR for the appearance of malignancies in the partially in-field organs after radiotherapy of male and female patients was (0.08%–1.8%) and (0.09%–1.9%), respectively. These risk values were 1.5–15.5 times lower when compared to the respective LIRs. The lifetime probability for out-of-field cancer induction in irradiated males and females was (2.5 × 10{sup −4} to 7.7 × 10{sup −2})% and (1.4 × 10{sup −4} to 2.6 × 10{sup −1})%, respectively. The above risks were one to four orders of magnitude lower than the LIRs. Conclusions: The probability for the development of out-of-field malignancies due to radiotherapy for VHs is trivial with respect to the nominal risk for unexposed population. The respective cancer risks to partially in-field organs are smaller than the nominal probabilities but they should not be considered as inconsiderable. These risks may be taken into account during the follow-up of patients treated for a symptomatic VH.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2022

Computational models of direct and indirect X‐ray breast imaging detectors for in silico trials

Abstract Background To facilitate in silico studies that investigate digital mammography (DM) and breast tomosynthesis (DBT), models replicating the variety in imaging performance of the DM and DBT systems, observed across manufacturers are needed. Purpose The main purpose of this work is to develop generic physics models for direct and indirect detector technology used in commercially available systems, with the goal of making them available open source to manufacturers to further tweak and develop the exact in silico replicas of their systems. Methods We recently reported on an in silico version of the SIEMENS Mammomat Inspiration DM/DBT system using an open‐source GPU‐accelerated Monte Carlo x‐ray imaging simulation code (MC‐GPU). We build on the previous version of the MC‐GPU codes to mimic the imaging performances of two other Food and Drug Administration (FDA)‐approved DM/DBT systems, such as Hologic Selenia Dimensions (HSD) and the General Electric Senographe Pristina (GSP) systems. In this work, we developed a hybrid technique to model the optical spread and signal crosstalk observed in the GSP and HSD systems. MC simulations are used to track each x‐ray photon till its first interaction within the x‐ray detector. On the other hand, the signal spread in the x‐ray detectors is modeled using previously developed analytical equations. This approach allows us to preserve the modeling accuracy offered by MC methods in the patient body, while speeding up secondary carrier transport (either electron–hole pairs or optical photons) using analytical equations in the detector. The analytical optical spread model for the indirect detector includes the depth‐dependent spread and collection of optical photons and relies on a pre‐computed set of point response functions that describe the optical spread as a function of depth. To understand the capabilities of the computational x‐ray detector models, we compared image quality metrics like modulation transfer function (MTF), normalized noise power spectrum (NNPS), and detective quantum efficiency (DQE), simulated with our models against measured data. Please note that the purpose of these comparisons with measured data would be to gauge if the model developed as part of this work could replicate commercially used direct and indirect technology in general and not to achieve perfect fits with measured data. Results We found that the simulated image quality metrics such as MTF, NNPS, and DQE were in reasonable agreement with experimental data. To demonstrate the imaging performance of the three DM/DBT systems, we integrated the detector models with the VICTRE pipeline and simulated DM images of a fatty breast model containing a spiculated mass and a calcium oxalate cluster. In general, we found that the images generated using the indirect model appeared more blurred with a different noise texture and contrast as compared to the systems with direct detectors. Conclusions We have presented computational models of three commercially available FDA‐approved DM/DBT systems, which implement both direct and indirect detector technology. The updated versions of the MC‐GPU codes that can be used to replicate three systems are available in open source format through GitHub.

Sengupta, Aunnasha↗
DOE OSTI2015

Top-pair production and decay at NLO matched with parton showers

We present a next-to-leading order (NLO) calculation of tt¯ production in hadronic collisions interfaced to shower generators according to the POWHEG method. We start from an NLO result from previous work, obtained in the zero width limit, where radiative corrections to both production and decays are included. The POWHEG interface required an extension of the POWHEG BOX framework, in order to deal with radiation from the decay of resonances. This extension is fully general (i.e. it can be applied in principle to any process considered in the zero width limit), and is here applied for the first time. In order to perform a realistic simulation, we introduce finite width effects using different approximations, that we validated by comparing with published exact NLO results. We have interfaced our POWHEG code to the PYTHIA8 shower Monte Carlo generator. At this stage, we dealt with novel issues related to the treatment of resonances, especially with regard to the initial scale for the shower that needs to be set appropriately. This procedure affects, for example, the fragmentation function of the b quark, that we have studied with particular attention. We believe that the tool presented here improves over previous generators for all aspects that have to do with top decays, and especially for the study of issues related to top mass measurements that involve B hadrons or b jets. As a result, the work presented here also constitutes a first step towards a fully consistent matching of NLO calculations involving intermediate resonances decaying into coloured particles, with parton showers.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2012

Nuclear spectroscopy for in situ soil elemental analysis: Monte Carlo simulations

We developed a model to simulate a novel inelastic neutron scattering (INS) system for in situ non-destructive analysis of soil using standard Monte Carlo Neutron Photon (MCNP5a) transport code. The volumes from which 90%, 95%, and 99% of the total signal are detected were estimated to be 0.23 m{sup 3}, 0.37 m{sup 3}, and 0.79 m{sup 3}, respectively. Similarly, we assessed the instrument's sampling footprint and depths. In addition we discuss the impact of the carbon's depth distribution on sampled depth.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2017

Spectroscopic and neutron detection properties of rare earth and titanium doped LiAlO 2 single crystals

In this study, LiAlO2 crystals doped with rare-earth elements and Ti were produced by the CZ method and spectroscopic and neutron detection properties were investigated. Photoluminescence revealed no clear luminescent activation of LiAlO2 by the rare-earth dopants though some interesting luminescence was observed from secondary phases within the crystal. Gamma-ray pulse height spectra collected using a 137Cs source exhibited only a Compton edge for the crystals. Neutron modeling using Monte Carlo N-Particle Transport Code revealed most neutrons used in the detection setup are thermalized, and while using natural lithium in the crystal growth, which contains 7.6 % 6Li, a 10 mm Ø by 10 mm sample of LiAlO2 has a 70.7 % intrinsic thermal neutron capture efficiency. Furthermore, the pulse height spectra collected using a 241Am-Be neutron source demonstrated a distinct neutron peak.

Dickens, Peter T.↗
DOE OSTI2021

Comparison of MCNP thermal scattering laws to inelastic neutron scattering data

Neutron transport computations employ thermal scattering laws (TSLs) to account for the influence of material structure and collective excitations at low energy (≲ 10 eV). Recent work has been undertaken to improve TSL accuracy and expand the number of available materials in the Evaluated Nuclear Data File (ENDF). However, comparison of TSL to measured data is often sparse. In this work, we present a method to evaluate TSL accuracy in the Monte Carlo N-Particle (MCNP) code by simulating inelastic neutron scattering (INS) measurement and comparing to data. The method is briefly described with example application to the recently released ENDF-VIII TSL library for polyethylene as well as a cryogenic PE TSL we are separately developing.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Monte Carlo N-Particle forward modeling for density reconstruction of double shell capsule radiographs

In the Double Shell Inertial Confinement Fusion concept, characterizing the shape asymmetry of imploding metal shells is vital for understanding energy-efficient compression and radiative losses of the thermonuclear fuel. The Monte Carlo N-Particle MCNP® code forward models radiography of Double Shell capsule implosions using the Advanced Radiographic Capability at the National Ignition Facility. A procedure is developed for using MCNP to reconstruct density profiles from the radiograph image intensity. For a given Double Shell imploding target geometry, MCNP radiographs predict image contrast, which can help guide experimental design. Finally, in future work, the calculated MCNP synthetic radiographs will be compared with experimental radiographs to determine the radial and azimuthal density profiles of the Double Shell capsules.

47 OTHER INSTRUMENTATION↗