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At least 1,117 records · Page 62
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Poster - 40: Treatment Verification of a 3D-printed Eye Phantom for Proton Therapy

Purpose: Ocular melanoma is a form of eye cancer which is often treated using proton therapy. The benefit of the steep proton dose gradient can only be leveraged for accurate patient eye alignment. A treatment-planning program was written to plan on a 3D-printed anatomical eye-phantom, which was then irradiated to demonstrate the feasibility of verifying in vivo dosimetry for proton therapy using PET imaging. Methods: A 3D CAD eye model with critical organs was designed and voxelized into the Monte-Carlo transport code FLUKA. Proton dose and PET isotope production were simulated for a treatment plan of a test tumour, generated by a 2D treatment-planning program developed using NumPy and proton range tables. Next, a plastic eye-phantom was 3D-printed from the CAD model, irradiated at the TRIUMF Proton Therapy facility, and imaged using a PET scanner. Results: The treatment-planning program prediction of the range setting and modulator wheel was verified in FLUKA to treat the tumour with at least 90% dose coverage for both tissue and plastic. An axial isotope distribution of the PET isotopes was simulated in FLUKA and converted to PET scan counts. Meanwhile, the 3D-printed eye-phantom successfully yielded a PET signal. Conclusions: The 2D treatment-planning program can predict required parameters to sufficiently treat an eye tumour, which was experimentally verified using commercial 3D-printing hardware to manufacture eye-phantoms. Comparison between the simulated and measured PET isotope distribution could provide a more realistic test of eye alignment, and a variation of the method using radiographic film is being developed.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2012

Use of freeze-casting in advanced burner reactor fuel design

This paper will detail the modeling of a fast reactor with fuel pins created using a freeze-casting process. Freeze-casting is a method of creating an inert scaffold within a fuel pin. The scaffold is created using a directional solidification process and results in open porosity for emplacement of fuel, with pores ranging in size from 300 microns to 500 microns in diameter. These pores allow multiple fuel types and enrichments to be loaded into one fuel pin. Also, each pore could be filled with varying amounts of fuel to allow for the specific volume of fission gases created by that fuel type. Currently fast reactors, including advanced burner reactors (ABR's), are not economically feasible due to the high cost of operating the reactors and of reprocessing the fuel. However, if the fuel could be very precisely placed, such as within a freeze-cast scaffold, this could increase fuel performance and result in a valid design with a much lower cost per megawatt. In addition to competitive costs, freeze-cast fuel would also allow for selective breeding or burning of actinides within specific locations in fast reactors. For example, fast flux peak locations could be utilized on a minute scale to target specific actinides for transmutation. Freeze-cast fuel is extremely flexible and has great potential in a variety of applications. This paper performs initial modeling of freeze-cast fuel, with the generic fast reactor parameters for this model based on EBR-II. The core has an assumed power of 62.5 MWt. The neutronics code used was Monte Carlo N-Particle (MCNP5) transport code. Uniform pore sizes were used in increments of 100 microns. Two different freeze-cast scaffold materials were used: ceramic (MgO-ZrO{sub 2}) and steel (SS316L). Separate models were needed for each material because the freeze-cast ceramic and metal scaffolds have different structural characteristics and overall porosities. Basic criticality results were compiled for the various models. Preliminary results show that criticality is achievable with freeze-cast fuel pins despite the significant amount of inert fuel matrix. Freeze casting is a promising method to achieve very precise fuel placement within fuel pins. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Calculations versus measurements for residual dose rates from SNS spent structures

Here, residual dose rate measurements were conducted on target vessel #13 and proton beam window #5 after extraction from their service locations. These measurements are used to verify calculation methods of radionuclide inventory assessment that are typically performed for nuclear waste characterization and transportation of these structures. Neutronics analyses for predicting residual dose rates are carried out using the transport code MCNPX and the transmutation code CINDER90. For transport analyses a complex and rigorous geometry model of the structures and their surroundings are applied. The neutronics analyses are carried out using the Bertini and CEM high energy physics models for simulating particles interactions above the table-based cross section range. Finally, obtained calculational results are analyzed and compared to the measured dose rates and overall show good agreement within 25%, which shows applicability of the methods used in analyses.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRSDate not supplied

ANALYSIS OF THE MSL/MEDLI ENTRY DATA WITH COUPLED CFD AND MATERIAL RESPONSE.

The Mars Science Laboratory (MSL) was protected during its atmospheric entry by an instrumented heat-shield using NASA's Phenolic Impregnated Carbon Ablator (PICA) material. PICA is a lightweight carbon fiber/polymeric resin material that offers out-standing performances for protecting probes during planetary entry. The Mars Entry Descent and Landing Instrument (MEDLI) suite on MSL offers unique in-flight validation data for models of material response and atmospheric entry. MEDLI recorded, among other things, time-resolved in-depth temperature data of PICA using thermocouple sensors assembled in the MEDLI Integrated Sensor Plugs (MISP). The objective of this work is to showcase and analyze the coupling between the material response and the aerothermal environment. As shown in Figure 1, the workflow is divided into the following steps. First, the aerothermal properties are computed in the Data Parallel Line Relaxation (DPLR) code [3] and used with the Nonequilibrium air radiation (NEQAIR) program [8] to compute radiative heating. Second, the thermal response inside the material is computed in the Porous material Analysis Toolbox based on Open-FOAM (PATO) using a fixed blowing correction parameter. Third, the pyrolysis gases computed in PATO are used as inputs to a blowing boundary condition within DPLR. Fourth, the new environment properties from DPLR are used in NEQAIR to provide an updated solution, then both the updated aerothermal environment and radiative heating are used in PATO without blowing correction. The third and fourth steps are then repeated until convergence in surface temperature is obtained. Convergence in the radiative heating is generally achieved before surface temperature, at which point the radiative heating is no longer updated. Char mass loss rates are forced to zero to produce a non-receding surface condition. For early time points in the trajectory, where flow around the MSL aeroshell is rarefied, the Direct Simulation Monte Carlo (DSMC) code, SPARTA, is used to compute the aerothermal environment. Iteration between PATO and SPARTA is not performed due to the computational cost of DSMC simulations. Preliminary results of the coupling between PATO and DPLR for the MSL heatshield atmospheric entry model are presented in Figures 2-4 at 65 seconds after entry interface. Figure 2 shows the surface temperature results from an uncoupled simulation in PATO with the blowing correction parameter applied (left) along with the coupled surface temperature after iteration (right). Figure 3 shows the surface temperature along the centerline from windward to leeward for easier comparison. Figure 4 shows the coupled and uncoupled pyrolysis gas blowing rate. Mars 2020 used a similar heatshield consisting of PICA for thermal protection during entry, descent, and landing. In preparation for Mars 2020 post-flight analysis, the predictive material response capability is benchmarked against flight data from MEDLI. This work represents an important milestone toward the development of validated predictive capabilities for designing thermal protection systems for planetary probes.

Mars Science Laboratory↗
NASA NTRSDate not supplied

Analysis of MSL/MEDLI Entry Data with Coupled CFD and Material Response

The Mars Science Laboratory (MSL) was protected during its atmospheric entry by an instrumented heatshield using NASA's Phenolic Impregnated Carbon Ablator (PICA) material [1]. PICA is a lightweight carbon fiber/polymeric resin material that offers outstanding performances for protecting probes during planetary entry. The Mars Entry Descent and Landing Instrument (MEDLI) suite on MSL offers unique in-flight validation data for models of material response and atmospheric entry. MEDLI recorded, among other things, time-resolved in-depth temperature data of PICA using thermocouple sensors assembled in the MEDLI Integrated Sensor Plugs (MISP) [2]. The objective of this work is to showcase and analyze the coupling between the material response and the aerothermal environment. As shown in Figure 1, the workflow is divided into the following steps. First, the aerothermal properties are computed in the Data Parallel Line Relaxation (DPLR) code [3] and used with the Nonequilibrium air radiation (NEQAIR) program [8] to compute radiative heating. Second, the thermal response inside the material is computed in the Porous material Analysis Toolbox based on OpenFOAM (PATO) [4,5,6] using a fixed blowing correction parameter. Third, the pyrolysis gases computed in PATO are used as inputs to a blowing boundary condition within DPLR. Fourth, the new environment properties from DPLR are used in NEQAIR to provide an updated solution, then both the updated aerothermal environment and radiative heating are used in PATO without blowing correction. The third and fourth steps are then repeated until convergence in surface temperature is obtained. Convergence in the radiative heating is generally achieved before surface temperature, at which point the radiative heating is no longer updated. Char mass loss rates are forced to zero to produce a non-receding surface condition. For early time points in the trajectory, where flow around the MSL aeroshell is rarefied, the Direct Simulation Monte Carlo (DSMC) code, SPARTA [7], is used to compute the aerothermal environment. Iteration between PATO and SPARTA is not performed due to the computational cost of DSMC simulations. Preliminary results of the coupling between PATO and DPLR for the MSL heatshield atmospheric entry model are presented in Figures 2-4 at 65 seconds after entry interface. Figure 2 shows the surface temperature results from an uncoupled simulation in PATO with the blowing correction parameter applied (left) along with the coupled surface temperature after iteration (right). Figure 3 shows the surface temperature along the centerline from windward to leeward for easier comparison. Figure 4 shows the coupled and uncoupled pyrolysis gas blowing rate. Mars 2020 used a similar heatshield consisting of PICA for thermal protection during entry, descent, and landing. In preparation for Mars 2020 post-flight analysis, the predictive material response capability is benchmarked against flight data from MEDLI. This work represents an important milestone toward the development of validated predictive capabilities for designing thermal protection systems for planetary probes.

Thermal Protection Systems↗
DOE OSTI2016

Effect of the size of nanoparticles on their dissolution within metal-glass nanocomposites under sustained irradiation

We investigate the dissolution law of metallic nanoparticles (NPs) under sustained irradiation. The system is composed of isolated spherical gold NPs (4–100 nm) embedded in an amorphous silica host matrix. Samples are irradiated at room temperature in the nuclear stopping power regime with 4 MeV Au ions for fluences up to 8 × 10{sup 16 }cm{sup −2}. Experimentally, the dependence of the dissolution kinetics on the irradiation fluence is linear for large NPs (45–100 nm) and exponential for small NPs (4–25 nm). A lattice-based kinetic Monte Carlo (KMC) code, which includes atomic diffusion and ballistic displacement events, is used to simulate the dynamical competition between irradiation effects and thermal healing. The KMC simulations allow for a qualitative description of the NP dissolution in two main stages, in good agreement with the experiment. Moreover, the perfect correlation obtained between the evolution of the simulated flux of ejected atoms and the dissolution rate in two stages implies that there exists an effect of the size of NPs on their dissolution and a critical size for the transition between the two stages. The Frost-Russell model providing an analytical solution for the dissolution rate, accounts well for the first dissolution stage but fails in reproducing the data for the second stage. An improved model obtained by including a size-dependent recoil generation rate permits fully describing the dissolution for any NP size. This proves, in particular, that the size effect on the generation rate is the principal reason for the existence of two regimes. Finally, our results also demonstrate that it is justified to use a unidirectional approximation to describe the dissolution of the NP under irradiation, because the solute concentration is particularly low in metal-glass nanocomposites.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2016

SU-G-JeP3-06: Lower KV Image Dose Are Expected From a Limited-Angle Intra-Fractional Verification (LIVE) System for SBRT Treatments

Purpose: In order to track the tumor movement for patient positioning verification during arc treatment delivery or in between 3D/IMRT beams for stereotactic body radiation therapy (SBRT), the limited-angle kV projections acquisition simultaneously during arc treatment delivery or in-between static treatment beams as the gantry moves to the next beam angle was proposed. The purpose of this study is to estimate additional imaging dose resulting from multiple tomosynthesis acquisitions in-between static treatment beams and to compare with that of a conventional kV-CBCT acquisition. Methods: kV imaging system integrated into Varian TrueBeam accelerators was modeled using EGSnrc Monte Carlo user code, BEAMnrc and DOSXYZnrc code was used in dose calculations. The simulated realistic kV beams from the Varian TrueBeam OBI 1.5 system were used to calculate dose to patient based on CT images. Organ doses were analyzed using DVHs. The imaging dose to patient resulting from realistic multiple tomosynthesis acquisitions with each 25–30 degree kV source rotation between 6 treatment beam gantry angles was studied. Results: For a typical lung SBRT treatment delivery much lower (20–50%) kV imaging doses from the sum of realistic six tomosynthesis acquisitions with each 25–30 degree x-ray source rotation between six treatment beam gantry angles were observed compared to that from a single CBCT image acquisition. Conclusion: This work indicates that the kV imaging in this proposed Limited-angle Intra-fractional Verification (LIVE) System for SBRT Treatments has a negligible imaging dose increase. It is worth to note that the MV imaging dose caused by MV projection acquisition in-between static beams in LIVE can be minimized by restricting the imaging to the target region and reducing the number of projections acquired. For arc treatments, MV imaging acquisition in LIVE does not add additional imaging dose as the MV images are acquired from treatment beams directly during the treatment.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2014

TU-F-BRE-03: Application of a Novel Mass-Density Compensation Optimization Method to Improve the Response of a Liquid-Filled Ionization Chamber in Nonstandard Fields

Purpose: To improve the response of the microLion detector (PTW 31018) in small field conditions by optimizing the density of the detector's non-sensitive components using a novel method based on the detector's dose response function. Methods: The central values h(0,0) of the perturbation functions for the microLion detector and a volume of water equivalent to its sensitive volume were calculated using the Monte Carlo user code egs-chamber by scoring the dose absorbed by a 6 MV photon pencil beam incident on their centroids. Values of h(0,0) were plotted as a function of the density of the microLion's graphite electrode with the detector placed in the axial orientation. The optimized density was found by finding the minimal value of h(0,0). Results: A density of 1.37 g/cm3 was found to minimize the perturbation function of the microLion detector. The modified microLion's response was then evaluated in small square fields with sides in the range of 5 – 40 mm and found to be consistent with highly watere-quivalent detectors such as a scintillating detector (Exradin W1) and a generic alanine detector in both axial and radial orientations. Conclusion: This work illustrates a novel method which can used to optimize the design of radiation detectors in small fields. This method should also work with other optimization parameters (e.g. thickness of electrode). Density-compensated detectors have the potential to eliminate the need to evaluate nonstandard field correction factors as described by the IAEA-AAPM formalism (Alfonso et al.) and simplify future dosimetry protocols for SRS/SBRT modalities. Finally, we also expect an improvement in the response of density-compensated detectors for composite IMRT fields.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

A Mercury Model of the Molly-G Fast Burst Reactor (FBR)

In order to design neutron irradiation experiments at a given nuclear reactor, one requires a representative neutron spectrum for the configuration of the reactor that will exist during the experiment. This usually implies modeling the operation of the reactor using a neutron transport code. Since the modeling of pulsed neutron reactors is quite complex, requiring coupled modeling of neutronics and thermo-structural response, the typical methodology employs analysis of the neutron spectrum which is obtained from a static keff calculation. While this time-independent Eigenfunction is not truly representative of that during a reactor pulse, we postulate that it is a reasonable representation for our purposes. The Nuclear Survivability (NS) Program at LLNL is considering fielding such experiments at the Molly-G (‘Molybdenum Godiva-II’) Fast Burst Reactor (FBR) at the White Sands Missile Range (WSMR). The goal of these experiments is to (a) test / proof / calibrate diagnostics, and (b) test the efficacy of neutron shield configurations, before fielding at other nuclear reactors, such as the Annular Core Research Reactor (ACRR) at SNL/NM. After attempting to secure either drawings or a copy of an existing MCNP model of the Molly-G FBR from various institutions, it was decided to develop a new model for use with the Mercury Monte Carlo transport code. The model described herein will be used to design these experiments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Exploring {sup 57}Co as a new isotope for brachytherapy applications

Purpose: The characteristics of the radionuclide {sup 57}Co make it interesting for use as a brachytherapy source. {sup 57}Co combines a possible high specific activity with the emission of relatively low-energy photons and a half-life (272 days) suitable for regular source exchanges in an afterloader. {sup 57}Co decays by electron capture to the stable {sup 57}Fe with emission of 136 and 122 keV photons. Methods: A hypothetical {sup 57}Co source based on the Flexisource brachytherapy encapsulation with the active core set as a pure cobalt cylinder (length 3.5 mm and diameter 0.6 mm) covered with a cylindrical stainless-steel capsule (length 5 mm and thickness 0.125 mm) was simulated using Geant4 Monte Carlo (MC) code version 9.4. The radial dose function, g(r), and anisotropy function F(r,{theta}), for the line source approximation were calculated following the TG-43U1 formalism. The results were compared to well-known {sup 192}Ir and {sup 125}I radionuclides, representing the higher and the lower energy end of brachytherapy, respectively. Results: The mean energy of photons in water, after passing through the core and the encapsulation material was 123 keV. This hypothetical {sup 57}Co source has an increasing g(r) due to multiple scatter of low-energy photons, which results in a more uniform dose distribution than {sup 192}Ir. Conclusions: {sup 57}Co has many advantages compared to {sup 192}Ir due to its low-energy gamma emissions without any electron contamination. {sup 57}Co has an increasing g(r) that results in a more uniform dose distribution than {sup 192}Ir due to its multiple scattered photons. The anisotropy of the {sup 57}Co source is comparable to that of {sup 192}Ir. Furthermore, {sup 57}Co has lower shielding requirements than {sup 192}Ir.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2014

Experiment Design of Secondary Neutron and Charged Particles Measurement with Stopping Targets Bombarded by 100- and 230-MeV/amu Helium Ions - Paper 124

Radiotherapy using ion beams has gained momentum in radiation oncology. While the number of proton therapy facilities that have been established or are under construction is rapidly growing, interest in heavy ion (whose atomic number Z is greater than one) radiotherapy has also been revitalized in the U.S. Helium ions, with mass and charge between protons and carbon ions, have advantages of deep penetration distances with significant peak-to-entrance dose ratios and limited dose contributions from secondary neutrons and other secondary charged fragments. In consideration of these advantages, the U.S. ion beam therapy community has expressed interests in helium ions as a potential candidate for ion beam therapy, especially pediatric cases. As such, characterizing the secondary nuclear fragment field created by primary helium ion interactions along the beam delivery devices and in human tissue is important for dose assessment and patient protection. Up to now the experimental data with respect to helium ions over the therapeutic energy range, 100 - 230 MeV/amu, is scarce compared to protons and carbon ions. Kurosawa et al. measured stopping target yields and angular spectra of neutrons, protons, deuterons and tritons produced by various target materials bombarded by 100- and 180- MeV/amu helium ions; Heilbronn et al. and Cecil et al. also measured the stopping target neutron yields by 155-, 160- and 177.5-MeV/amu helium ion bombardments. We have proposed to measure angular distributions of secondary neutrons and light charged ions produced by 100- and 230-MeV/amu helium ions bombarding stopping water and iron targets. The experiment was conducted at the Heavy Ion Medical Accelerator in Chiba (HIMAC) of the National Institute of Radiological Sciences (NIRS) in Japan this June. During the literature survey for experimental setup, we found that the ratio of the ion beam range (R) to the target depth (d) varies from 0.5 - 0.98 in the stopping target measurements either for helium ions incident on various targets or carbon ions bombarding targets as tissue surrogates. The difference in the values of ratio R/d makes comparison of experimental results complicated and require accurate corrections. Hence, how the R/d ratio and cross sectional area of the targets influence the angular distributions of secondary particles is investigated by using a Monte Carlo method in this study. The influence of the depth and the cross sectional area of stopping water and iron targets on the yield and the spectra of secondary particles induced by 100- and 230-MeV/amu He ions was studied by a Monte Carlo simulation code, PHITS. The secondary charged particles and secondary neutrons who come from breakup of projectiles are highly forward-focused, while the secondary particles produced by target breakup are nearly isotropic. As the energy of {sup 4}He ions increases from 100 MeV/amu to 230 MeV/amu, the normalized intensities of charged particles decrease more gently with increasing angles, whereas the neutron angular distribution falls quickly at angles smaller than 30 deg. and starts to exceed the normalized profile of 100 MeV/amu He from 60 deg. When the target depth increases, the impact on attenuation of secondary particles is more significant for lighter target mass and higher-energy projectiles at forward angles. Also, with deeper targets, more interactions occur between the secondary particles and the target elements, which results in more target-like fragments at large and backward angles. In respect of the cross sectional area of the stopping targets, the forward angular distributions are similar to the system with smaller cross sectional area of the targets; however, no general rule was found for how the large-angle distributions are affected by the target cross sectional area. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

Accelerated Coupled Monte Carlo-Thermal Hydraulic Calculations using a Hybrid GTF-Diffusion-based Prediction Block: First Results

Accurate predictions of spatial power and temperature distributions require the coupling of a neutron transport solver with a thermal-hydraulic (TH) feedback. Nowadays, Monte Carlo (MC) codes are widely coupled to TH solvers, typically via a Picard iteration (PI) method, due to the higher fidelity that such frameworks can produce. To speed up a PI, a prediction step can produce an improved initial guess for a source distribution and feed it to the MC code. Recent work [1, 2] investigated a prediction step that uses generalized transfer functions (GTFs) to predict the macroscopic cross sections’ variations following a perturbation in TH properties, such as coolant density. The previous method also relied on first order perturbation (FOP) theory to predict perturbed power profiles, rather than using an expensive MC iterate. The implemented FOP method relied on generating a fission matrix from which the forward and adjoint eigenmodes were extracted and later used to for power calculations. The generation of the fission matrix can introduce a significant computational overhead, therefore undermining the performance of the proposed hybrid technique when applied to high-dimensional problems, e.g., full core calculations. This work attempts to improve the GTF-FOP prediction step by replacing the FOP solver with a nodal diffusion solver, thus eliminating the need to calculate a fission matrix. The GTF-diffusion step was tested for various moderator density perturbations. In each case, the predicted power distribution showed good agreement with the reference case. The latter is attributed to the generally good prediction of most spatially distributed macroscopic cross sections, except the transport cross section, which will become the focus of future work.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Prediction of neutron production and energy spectrum by the inverse kinematic reaction between an incident 7 Li 3+ beam and a proton target in PHITS

A neutron source using the inverse kinematic reaction between lithium and proton, p( 7 Li, n) 7 Be, achieves forward-directed neutrons, potentially enhancing neutron yield in the forward direction. Despite the advantage, no evaluated-cross-section data for this reaction can be used in Monte Carlo simulation codes, such as PHITS. To solve this problem, this study aims to evaluate the applicability of the user-defined cross-section data, Frag data, for p ( 7 Li, n) 7 Be in PHITS. The simulations reproduced collisions between 7 Li 3+ ions and polypropylene targets. The Frag data was edited based on the JENDL-5 by utilizing the two-body collision kinematics. The neutron yield and angular distribution were investigated in the simulation. As a result, the forward neutron convergence with a reasonable neutron yield and energy spectrum was observed. The expected neutron yield in the forward 1-steradian area is 2.46 × 10 10 n/s when lithium-ion energy and current are 16.45 MeV and 0.1 mA.

43 PARTICLE ACCELERATORS↗
DOE OSTI2012

Radiation Field of Packages Carrying Spent Co-60 Radioactive Sources - 12437

Among the diverse radioactive sources commonly exploited in medical and industrial applications, Co- 60 is increasingly used as strong gamma emitter. Over time, source manufacturers favored Co-60 as opposed to other gamma emitters because its relatively short half-life (5.27 year) that minimizes issues related to the management of disused sources. Disused Co-60 sources can retain a significant amount of radioactivity (from hundreds of MBq to several GBq) that still poses safety concerns on their handling and transportation. In this context a detailed knowledge of their radiation field would provide the necessary information for taking actions in preventing unnecessary doses to the workers and the population by optimizing transportation procedures and handling operations. We modeled the geometry and the materials constituting a transportation packaging of a spent Co-60 source which had an original maximum activity of a few GBq and was enclosed in a small lead irradiator. Then we applied a Monte Carlo transport code (MCNP5) for tracking down the gamma photons emitted by the source, including the secondary photons resulting by the interaction of the source photons with the surrounding materials. This allowed for the evaluation of the radiation field inside and outside the packaging, and the corresponding equivalent dose useful for checking the compliance with the regulations and the health risk of possible radiation exposure. We found that a typical 60-liters drum carrying a spent Co-60 source, enclosed in its original irradiator, with a residual activity of 300 MBq could already overcome an equivalent dose of 0.2 mSv/h on the drum external surface, which is the maximum equivalent dose at any point of the surface for this packaging as prescribed by local regulations. This condition is even more apparent when the source is slightly displaced with respect to the rotation axis of the drum, an easily occurring condition for sources not properly packaged, generating non-compliant hot-spots on the drum surface. As an example, a displacement of 5 cm translates in an increase of 80% in the dose level on the nearest side of the drum. We also found that the equivalent dose is significantly influenced by the scattered source photons and the secondary photons, whose contribution to the radiation field is mainly determined by the package geometry. The developed model resulted in an important tool for exploring the detail of the radiation field of a spent Co-60 source packaged for transportation allowing to check for compliance with the regulations and to evaluate risks to the workers and the population. It is worth to point out that this modeling approach is completely general and can be applied to a variety of different problems not limited to the transportation of radioactive material. In the example illustrated in this work, a small Co-60 source inside its irradiator and properly packaged in a drum for transport, presents a hot-spot on the drum surface if the irradiator is displaced of just few centimeters with respect to the rotation axis of the drum. A Monte Carlo simulation provided a complete description of the radiation field in the volume surrounding the packaging as a function of photon energy showing that a significant contribution comes from source photons scattered by the package itself. As an example, the Transport Index (TI) of this packaging was calculated, resulting in a value of 0.5. This value is low when compared to the regulations, however suggests that an individual just 1 meter afar from the packaging receives, in a few hours, a dose comparable to that received with a radiograph. This might be the case of the driver of a truck transporting the packaging, in particular when a number of packages are transported at the same time. It is important to notice that the results presented in this work are general, and can be appropriately scaled to represent the radiation field of similar geometries with gamma sources of different activity. However, this quick application of a Monte Carlo simulation provides useful insights in the process of preparing a packaging for transport. This simple model suggests that this approach can be efficient for screening a number of packaging configurations and flagging those potentially not compliant with regulations, that will require a more thorough care in the packaging process. Furthermore the knowledge of the details of the radiation field can be useful for optimizing the geometry of the transport cargo, and becomes essential when a variety of packaging are transported or stored, in order to prevent hazardous situations during their handling, and hence preventing unnecessary doses to the workers and the populations. It is worth to point out that this modeling approach is completely general and can be applied to a variety of different problems not limited to the transportation of radioactive material. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2013

Numerical study of atomic production rate in hydrogen negative ion sources with the effect of non-equilibrium electron energy distribution function

Spatial non-uniformity of the dissociative hydrogen atom (H{sup 0}) production has been investigated in a large negative ion source (JAEA 10 A source) with the electron energy distribution function (EEDF) obtained by a Monte-Carlo simulation code for electron transport in 3D3V (three dimensional real and velocity) space. It has been shown that the H{sup 0} production rate becomes larger in the upper region (one side in the longitudinal direction) of the source chamber. This spatial non-uniformity of the H{sup 0} production profile is mainly explained by the non-equilibrium features of the EEDF in the upper region, i.e., the EEDF consists of thermal electron component with kinetic energy ε < 25 eV and fast electron component with energy ε > 25 eV in the upper region, while the EEDF mainly consists of only thermal electrons in the bottom region. These characteristics for the EEDF and the energy dependence of cross-sections for dissociation and dissociative ionization processes lead to the non-uniform profile of the H{sup 0} production. The above numerical results of the spatial H{sup 0} non-uniformity are validated and confirmed by comparisons with those by spectroscopic measurement. It has been clarified that the non-equilibrium (fast electron) component of the EEDF has a large contribution to the non-uniformity of the H{sup 0} production rate.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2014

SU-E-T-117: Dose to Organs Outside of CT Scan Range- Monte Carlo and Hybrid Phantom Approach

Purpose: Epidemiological study of second cancer risk for cancer survivors often requires the dose to normal tissues located outside the anatomy covered by radiological imaging, which is usually limited to tumor and organs at risk. We have investigated the feasibility of using whole body computational human phantoms for estimating out-of-field organ doses for patients treated by Intensity Modulated Radiation Therapy (IMRT). Methods: Identical 7-field IMRT prostate plans were performed using X-ray Voxel Monte Carlo (XVMC), a radiotherapy-specific Monte Carlo transport code, on the computed tomography (CT) images of the torso of an adult male patient (175 cm height, 66 kg weight) and an adult male hybrid computational phantom with the equivalent body size. Dose to the liver, right lung, and left lung were calculated and compared. Results: Considerable differences are seen between the doses calculated by XVMC for the patient CT and the hybrid phantom. One major contributing factor is the treatment method, deep inspiration breath hold (DIBH), used for this patient. This leads to significant differences in the organ position relative to the treatment isocenter. The transverse distances from the treatment isocenter to the inferior border of the liver, left lung, and right lung are 19.5cm, 29.5cm, and 30.0cm, respectively for the patient CT, compared with 24.3cm, 36.6cm, and 39.1cm, respectively, for the hybrid phantom. When corrected for the distance, the mean doses calculated using the hybrid phantom are within 28% of those calculated using the patient CT. Conclusion: This study showed that mean dose to the organs located in the missing CT coverage can be reconstructed by using whole body computational human phantoms within reasonable dosimetric uncertainty, however appropriate corrections may be necessary if the patient is treated with a technique that will significantly deform the size or location of the organs relative to the hybrid phantom.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2025

X-Ray Spectra from General Relativistic Radiation Magnetohydrodynamic Simulations of Thin Disks

We compare X-ray emission from several general relativistic, multifrequency, radiation magnetohydrodynamic simulations of thin black hole accretion disks with different accretion rates and spins. The simulations were performed using the M1 closure scheme, resolved with 12 frequency (energy) bins logarithmically spaced from 5 × 10 −3 to 5 × 10 3 ​​​ keV. We apply a general relativistic Monte Carlo transport code to postprocess the simulation data with greater fidelity in frequency resolution and Compton scattering treatment. Despite the relatively few energy bins and Kompaneets approximation to Compton scattering utilized in the M1 method, we find generally good agreement between the methods. Both produce prominent thermal profiles with peaks around 2–2.5 keV, where agreement is particularly strong and representative of the soft state. Both also find weaker (lower-luminosity) thermally sourced emission extending out to 100 keV due to the hotter innermost regions of the disks. Inverse Compton scattering becomes increasingly effective at hardening spectral outputs with increasing black hole spin, and becomes the dominant mechanism for photons that escape with energies between 10 to several hundred keV. At very high rates of spin, the radiation flux in this upscattered component becomes comparable to the thermal flux, a phenomenon typically associated with intermediate states. Beyond 10 4 keV, we observe faint, free–free emission from hot, optically thin coronal regions developing near the horizon, common to both spinning and nonspinning black holes.

79 ASTRONOMY AND ASTROPHYSICS↗