Monte Carlo simulation technology based on characteristic γ-ray spectrum analysis

Analysis of inconsistent source sampling in monte carlo weight-window variance reduction methods

The application of the Automated Variance Reduction Generator (ADVANTG) code to accelerate MCNP neutron transport calculations in fusion-relevant geometries is presented. The ADVANTG code generates variance-reduction parameters using the Consistent Adjoint Driven Importance Sampling (CADIS) and Forward-Weighted Consistent Adjoint Driven Importance Sampling (FW-CADIS) methods based on deterministic transport calculations performed by the discrete ordinates code Denovo. The aim of ADVANTG is to reduce the MCNP computational time by automating the process of variance-reduction parameter generation. ADVANTG was tested on a simplified model of a JET-like tokamak that in spite of its simplicity retains all the major characteristics of such a tokamak. The performance of the nuclear data libraries provided with ADVANTG and of various other ADVANTG/Denovo settings on variance-reduction efficiency was tested. Several cases using deuterium-deuterium or deuterium-tritium (D-T) volumetric (plasma) sources and 252Cf or D-T point neutron sources were analyzed to find guidelines for successful use of the code for fusion applications. Additionally, the use of ADVANTG as a tool to identify major neutron pathways from the neutron source to the detector is demonstrated.

To characterize an ionization chamber BEV‐CC01 as a standard of absorbed dose to water Dw at SSDL‐Mexico, the approach developed by the BIPM for 60Co gamma radiation, [1] has been chosen. This requires the estimation of a factor kp, which stems from the perturbation introduced by the presence of the ionization chamber in the water phantom, and due to finite size of the cavity. This factor is the product of four terms: ψw,c, (μenρ)w,c, (1 + μ′.ȳ)w,c and kcav. Two independent determinations are accomplished using a combination of the Monte Carlo code MCNP4C in ITS mode [2,3] and analytic methods: one kp∥=1.1626 ± uc=: 0.90% for the chamber axis parallel to the beam axis; and another kp =1.1079± uc=0.89% for the chamber axis perpendicular to the beam axis. The variance reduction techniques: splitting‐Russian roulette, source biasing and forced photon collisions are employed in the simulations to improve the calculation efficiency. The energy fluence for the 60Co housing‐source Picker C/9 is obtained by realistic Monte Carlo (MC) simulation, it is verified by comparison of MC calculated and measured beam output air kerma factors, and percent depth dose curves in water, PDD. This spectrum is considered as input energy for a point source (74% is from primary photons and the rest 26% is from scattered radiation) in the determination of the kp factors. Details of the calculations are given together with the theoretical basis of the ionometric standard employed.To characterize an ionization chamber BEV‐CC01 as a standard of absorbed dose to water Dw at SSDL‐Mexico, the approach developed by the BIPM for 60Co gamma radiation, [1] has been chosen. This requires the estimation of a factor kp, which stems from the perturbation introduced by the presence of the ionization chamber in the water phantom, and due to finite size of the cavity. This factor is the product of four terms: ψw,c, (μenρ)w,c, (1 + μ′.ȳ)w,c and kcav. Two independent determinations are accomplished using a combination of the Monte Carlo code MCNP4C in ITS mode [2,3] and analytic methods: one kp∥=1.1626 ± uc=: 0.90% for the chamber axis parallel to the beam axis; and another kp =1.1079± uc=0.89% for the chamber axis perpendicular to the beam axis. The variance reduction techniques: splitting‐Russian roulette, source biasing and forced photon collisions are employed in the simulations to improve the calculation efficiency. The energy fluence for the 60Co housing‐source Picker C/9 is obtained by reali...

Radiation dose rate distributions of spent fuel dry casks estimated with MAVRIC based on detailed geometry and continuous-energy models

Monte Carlo (MC) simulations are often used in calculations of radiation transport to enable accurate prediction of radiation-dose, even though the computation is relatively time-consuming. In a typical MC simulation, significant computation time is allocated to following non-important events. To address this issue, variance reduction techniques (VRTs) have been suggested for reducing the statistical variance for the same computation time. Among the available MC simulation codes, electron gamma shower (National Research Council of Canada) (EGSnrc) is a general-purpose coupled electron-photon transport code that also features an even-handed, rich set of VRTs. The most well-known VRTs are the photon splitting, Russian roulette (RR), and photon cross-section enhancement (XCSE) techniques. The objective of this work was to determine the optimal combination of VRTs that increases the simulation speed and the efficiency of simulation, without compromising its accuracy. Selection of VRTs was performed using EGSnrc MC User codes, such as cavity and egs_chamber, for simulating various ion chamber geometries using 6 MV photon beams and 1.25 MeV 60Co photon beams. The results show that the combination of XCSE and RR yields the highest efficiency for ion-chamber dose calculations inside a 30 cm × 30 cm × 30 cm water phantom. Hence, properly selecting a different VRT without altering the underlying physics increases the efficiency of MC simulations for ion-chamber dose calculation.

Production of secondary electrons and photons and energy absorption mechanisms in amorphous carbon irradiated by photons in the energy range of 0.03 to 17.4 keV

PurposeTo explain the deviation observed between measured and Monaco calculated dose profiles for a small field (i.e., alternating open‐closed MLC pattern). A Monte Carlo (MC) model of an Elekta Infinity linac with Agility MLC was created and validated against measurements. In addition, an analytic model which predicts the fluence at the isocenter plane was used to study the impact of multiple beam parameters on the accuracy of dose calculations for small fields.MethodsA detailed MC model of a 6 MV Elekta Infinity linac with Agility MLC was created in EGSnrc/BEAMnrc and validated against measurements. An analytic model using primary and secondary virtual photon sources was created and benchmarked against the MC simulations and the impact of multiple beam parameters on the accuracy of the model for a small field was investigated. Both models were used to explain discrepancies observed between measured/EGSnrc simulated and Monaco calculated dose profiles for alternating open‐closed MLC leaves.ResultsMC‐simulated dose profiles (PDDs, cross‐ and in‐line profiles, etc.) were found to be in very good agreements with measurements. The best fit for the leaf bank rotation was found to be 9 mrad to model the defocusing of Agility MLC. Moreover, a very good agreement was observed between results from the analytic model and MC simulations for a small field. Modifying the radial size of the incident electron beam in the BEAMnrc model improved the agreement between Monaco and EGSnrc calculated dose profiles by approximately 16% and 30% in the position of maxima and minima, respectively.ConclusionAccurate modeling of the full‐width‐half‐maximum (FWHM) of the primary photon source as well as the MLC leaf design (leaf bank rotation, etc.) is essential for accurate calculations of dose delivered by small radiation fields when using virtual source or MC models of the beam.

The ionizing radiation exposure to crew on current and future space missions can significantly increase their health risks for cancers, degenerative diseases, and other acute and late effects. A common approach for estimating risk to crew is by completing stochastic (e.g., Monte Carlo) or deterministic particle transport simulations. Within the simulated environment, a small fraction of the particle histories tracked will interact with the astronaut or detector, particularly for larger spacecraft such as the International Space Station, Tiangong Space Station or Lunar Gateway. These simulations can be computationally intensive as they require a very large number of particle histories to achieve a low statistical uncertainty. Variance reduction techniques are applied to simulations to reduce the computational time of the simulation while maintaining the same (or less) statistical uncertainty. The variance reduction technique developed herein involves applying a directional source bias to an isotropic radiation field, such as galactic cosmic rays, to reduce the quantity of particles that have a low probability of interacting with the astronaut or detector. A custom application has been developed utilizing the Geant4 Toolkit that computes the trajectories and energies of particles in three dimensions in the International Space Station using the Monte Carlo method. The results demonstrate the impact of our variance reduction technique on effective dose equivalence depending on: primary and secondary particle type (proton, neutron, photon, heavy ion, etc.), geometric volumes and spacecraft materials. Our variance reduction technique can be tuned by the user to optimize the simulation time depending on their objectives and enables rapid testing of different shield configurations and materials. This variance reduction technique is implemented easily using several input parameters for boundary conditions. Recommended values are presented for rapid implementation in simulations.

The capture cross sections and capture γ-ray spectra of 91,92Zr were measured in an incident neutron energy region from 15 to 100keV and at 550keV. A neutron time-of-flight method was adopted with a 1.5-ns pulsed neutron source by the 7Li(p,n)7Be reaction and with a large anti-Compton NaI(Tl) spectrometer. A pulse-height weighting technique was applied to observed capture γ-ray pulse-height spectra to obtain capture yields. Using the standard capture cross sections of 197Au, the capture cross sections of 91,92Zr were derived with the errors from 5 to 9%. The present results were compared with other measurements and the evaluated values in JENDL-3.3 and ENDF/B-VI. The JENDL-3.3 evaluations for 91Zr were in agreement with the present results, but those for 92Zr largely deviated from the present results. The capture γ-ray spectra of 91,92Zr were obtained by unfolding the observed capture γ-ray pulse-height spectra. The multiplicities of observed γ rays were derived from the γ-ray spectra.

Purpose: To develop a fast Monte Carlo code based on EGSnrc for accurate dose calculation around brachytherapy sources. Method and Materials: Sources and phantom geometries are modeled by using the Multi‐geometry technique which allows various predefined geometry elements (eg, sources, applicators, catheters) a phantom geometry. Sources such as an HDR Ir‐192 source and LDR I‐125 or Pd‐103 seed sources were modeled. One or more sources from a database can be duplicated many times and placed in arbitrary locations. Besides the above sources, BrachyDose can calculate dose around a miniature x‐ray‐tube source since it is based on EGSnrc. It also can use CT data in the phantom geometry. Variance reduction techniques are applied to speed up computation time. Dose is calculated by scoring the collision kerma using a tracklength estimator. There is an option to reuse every photon which escapes from a seed as if it came from every seed in the implant with same direction relative to the seed itself. Results: The speed of the BrachyDose calculation is specified by the time required to attain an average of 2\% statistics in the central region of an implant of 125 seeds spaced at 5 mm separation in a 1000 cm∧3 cubic phantom. The time required scales roughly as the inverse of the volume of the voxels. On an 2.4 MHz CPU, the computation time is 510 s for 1 mm voxels. The DVHs for 1 mm voxels are significantly different from those in 2 mm voxels. Changing phantom material from water to prostate tissue causes the dose to vary by +/−5% vs dose to water, depending on distance from the seeds. Conclusions: The Monte Carlo code developed is fast enough for routine clinical applications. Calculated dose values include inter‐seed effects and other effects from tissue inhomogeneities.

A cascaded linear system model that includes incomplete charge collection and interaction-depth dependent conversion gain and charge collection stages is considered for the calculation of the zero spatial frequency detective quantum efficiency, DQE(0), of a direct conversion x-ray image detector. The model includes signal and noise propagations in the following stages: (1) x-ray attenuation, (2) conversion gain, (3) charge collection, and (4) addition of electronic noise. The primary x-ray photon interaction and also the secondary K-fluorescent photon interaction are included in determining the interaction-depth dependent conversion gain across the photoconductor. We examine DQE(0) of a-Se detectors for fluoroscopic applications as a function of photoconductor thickness with varying amounts of electronic noise and x-ray exposure under (a) constant field, and (b) constant voltage operating conditions. We show that there is an optimum photoconductor thickness, which maximizes DQE(0) under a constant voltage operation. The optimum thickness depends on the added electronic noise, x-ray exposure, charge collection efficiency and bias voltage. For the quantities mentioned above that are appropriate for a-Se detectors and fluoroscopic applications, the optimum a-Se thickness is ∼700 μm and the corresponding DQE is ∼0.4. It is shown that the DQE depends strongly on the charge transport properties of the photoconductors. With the radiation-receiving electrode negatively biased, the DQE is more dependent on electron lifetime (τe) than hole lifetime (τh). Full electron trapping, (τe = 0) reduces the DQE by about 73.3% at the detector thickness of 1000 μm whereas full hole trapping (τh = 0) reduces the DQE by about 43.7%. The DQE for the negative bias is lower than for the positive bias, and the difference in DQE, as expected, increases with the photoconductor thickness because of the asymmetric transport properties of holes and electrons in a-Se. The present results show that the DQE generally does not continue to improve with greater photoconductor thickness because of charge carrier trapping effects. The DQE of a polyenergetic x-ray beam is only slightly lower than a monoenergetic x-ray beam with the same average photon energy. The theoretical model shows a very good agreement with the experimental DQE versus exposure characteristics published in the literature.

Godiva IV is a system comprised of highly enriched uranium alloyed with molybdenum in the form of fuel plate rings. The reactor, along with its predecessors, was designed with the unique ability to satisfy interests in the super-prompt-critical reactor operation space. Originally, the reactor was part of the Los Alamos Critical Experiments Facility (LACEF) at Technical Area-18 (TA-18). The radiation field around Godiva at this facility was well characterized and understood. As a fast neutron system, the neutron spectrum in and around Godiva was close to a Watt Fission spectrum. The Kiva where Godiva IV was located at LACEF was made of thin, sheet metal walls which did not contribute significantly to the neutron spectrum. Following the transition of LACEF to the National Critical Experiments and Research Center (NCERC) in Nevada, Godiva-IV was moved from TA-18 to the Device Assembly Facility (DAF) at the Nevada National Security Site (NNSS). Part of this move brought renewed interest in radiation field characterization. The new facility introduced significant changes to the environment surrounding Godiva, and preliminary foil irradiation results suggested that the room contribution to the neutron spectrum was significant. Unlike at TA-18, a large thermal neutron signature was added to the fast spectrum from Godiva due to significant room return. A primary goal due to the additional complexity that the room return adds to the Godiva IV radiation emission spectrum was the development of an efficient Monte Carlo N-Particle (MCNP) calculation capable of characterizing the neutron spectrum anywhere in the room around Godiva. A campaign of activation foil irradiations and analysis were completed to support the validation of the MCNP model. The modeling of these foils in MCNP can be easily done with a standard volumetric neutron flux tally. However, given the multitude of locations and reaction rates to be modeled, further steps must be taken to increase the efficiency of these calculations in MCNP. During this study, a benchmark model currently under development for Godiva IV was used. A qualitative assessment of the thermal neutron contributors was performed using spatial neutron distribution plots. Additional detail was added to the model based on the qualitative results showing the thermal spectrum’s large sensitivity to hydrogenous material. Neutron energy spectra was evaluated at discrete locations in the room around Godiva to quantify the relative contribution of various components. It was discovered that the concrete walls are the largest contributor to the thermal signature, with minor contributions from plastic components surrounding Godiva. Following these results, two different variance reduction techniques were implemented to improve the problem efficiency in these calculations. In the first approach, an F5 point detector tally was implemented in the standard Godiva IV criticality problem. The second approach involved a weight-window generator implementation with an F5 point detector tally in a fixed source problem. The weight window implementation reduced the runtime from 42739.55 minutes to 1803.34 minutes (computer time), compared to the F5 KCODE implementation.

Monte Carlo simulations are widely used in medical physics to model particle interactions for accurate radiotherapy dose calculations. This study presents a methodological comparison between two leading Geant4-based simulation platforms-GATE and TOPAS-for modeling a 6 MV Varian CLINAC IX linear accelerator. The objective was to evaluate and optimize the performance of both platforms in terms of computational efficiency and dosimetric accuracy. Simulations were validated against experimental measurements using percentage depth dose and dose profiles acquired in a water phantom. Methods included geometry modeling, electron source parameter tuning, and implementation of phase space files, production thresholds, and variance reduction techniques. Results showed that TOPAS provided superior agreement with experimental data, especially in deep dose regions, while GATE offered marginally better computation times when advanced variance reduction tools were applied. These findings highlight the importance of aligning simulation platform choice with specific clinical or research objectives. This work offers a practical reference for medical physicists and researchers seeking to optimize Monte Carlo simulations for treatment planning, quality assurance, or commissioning of radiotherapy systems.

Improved neutron inelastic scattering cross sections have repeatedly been identified as a top priority nuclear data need, important for basic science and a range of applications in nuclear energy, stockpile stewardship, and proliferation detection. For the C12(n,n′γ) reaction in particular, recent measurements have unveiled some structural discrepancies, demonstrating incongruities among themselves and in relation to the ENDF/B-VIII.0 nuclear data evaluation. To help resolve these disagreements, a measurement was performed at the 88-Inch Cyclotron at Lawrence Berkeley National Laboratory using a broad-spectrum neutron beam and a 99.8% pure natural carbon target. The Gamma Energy Neutron Energy Spectrometer for Inelastic Scattering (GENESIS) was employed to measure energy-differential γ-ray emission spectra as a function of incident neutron energy in the energy range of 5.5 to 16.7 MeV. The C12 partial γ-ray cross sections were extracted at 63∘, 122.5∘, and 150∘ with respect to the incoming neutron beam and integrated using angular distribution data available in the literature. The data show agreement with a recent literature measurement and evaluation from 11 to 15 MeV, but indicate a larger cross section for incident neutron energies between 5.5 and 8.5 MeV. The measured relative angular distributions are also reported and were found to agree with evaluation.

In this study, the absorbed dose was calculated to the small intestine (SI) wall of an adult human from electrons in its lumen contents. The effects on dose due to variations in the lumen radius and wall-thickness also were studied. The SI model was based on values gleaned from anatomic and histologic reviews of the adult human SI. Histologic and radiological analyses of the SI suggested the microscopic intricacy of this walled organ could be avoided for dosimetric purposes and a set of concentric cylinders could be used to model the SI. The model was input into the Monte Carlo N-Particle (MCNP) version 4A computational package, which was used to simulate energy deposition in the SI by electrons of fifty discrete energies ranging 10-500 keV. The source electrons as well as all resulting particles, such as knock-on electrons, bremsstrahlung, and electrons created from bremsstrahlung interactions, were transported until the particle energies fell below the 1 keV low-energy cutoff. Detailed physics treatments for secondary photons were made. With a reasonable number of histories, appropriate variance reduction techniques were used to improve the precision of the Monte Carlo calculations. The model used very small tally regions, which ranged in thickness from 0.5 microm to 200 microm depending on the electron energy studied and tally location in the wall. Relative errors associated with these calculations were maintained at less than 5%. The large number of tally results across the wall for each of the energies studied enabled the construction of the energy-specific depth dose curves in the wall. Each of these curves was consistent with the anticipated energy deposition pattern. These curves showed that only a small fraction of the energy absorbed at the contents-mucus interface reaches the stem cell layers because the cells are located deep in the mucosa. This fraction was found to vary from 1.66 x 10(-6) to 1.21 x 10(-1) over the energy range 10-500 keV. These results demonstrated the interface dose, which has been routinely reported as the "wall" dose, is a significant overestimate of the actual dose to the stem cells. The dose uncertainties associated with variations of the critical cell depth were shown to be very high for electrons whose CSDA ranges in the soft tissue exceeded the depth of the critical cells. This study showed that the uncertainty in the wall-thickness had no effect on depth doses while variation in the lumen radius significantly changes depth doses. The results suggest that these changes could be approximated by the inverse square of the lumen radius.