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At least 1,189 records · Page 66
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Preclinical investigation for developing injectable fiducial markers using a mixture of BaSO{sub 4} and biodegradable polymer for proton therapy

Purpose: The aim of this study is to investigate the use of mixture of BaSO{sub 4} and biodegradable polymer as an injectable nonmetallic fiducial marker to reduce artifacts in x-ray images, decrease the absorbed dose distortion in proton therapy, and replace permanent metal markers. Methods: Two samples were made with 90 wt. % polymer phosphate buffer saline (PBS) and 10 wt. % BaSO{sub 4} (B1) or 20 wt. % BaSO{sub 4} (B2). Two animal models (mice and rats) were used. To test the injectability and in vivo gelation, a volume of 200 μl at a pH 5.8 were injected into the Sprague-Dawley rats. After sacrificing the rats over time, the authors checked the gel morphology. Detectability of the markers in the x-ray images was tested for two sizes (diameters of 1 and 2 mm) for B1 and B2. Four samples were injected into BALB/C mice. The polymer mixed with BaSO{sub 4} transform from SOL at 20 °C with a pH of 6.0 to GEL in the living body at 37 °C with a pH of 7.4, so the size of the fiducial marker could be controlled by adjusting the injected volume. The detectability of the BaSO{sub 4} marker was measured in x-ray images of cone beam CT (CBCT), on-board imager [anterior–posterior (AP), lateral], and fluoroscopy (AP, lateral) using a Novalis-TX (Varian Medical Systems, Palo Alto, CA) repeatedly over 4 months. The volume, HU, and artifacts for the markers were measured in the CBCT images. Artifacts were compared to those of gold marker by analyzing the HU distribution. The dose distortion in proton therapy was computed by using a Monte Carlo (MC) code. A cylindrical shaped marker (diameter: 1 or 2 mm, length: 3 mm) made of gold, stainless-steel [304], titanium, and 20 wt. % BaSO{sub 4} was positioned at the center of the spread-out Bragg peak (SOBP) in parallel or perpendicular to the beam entrance. The dose distortion was measured on the depth dose profile across the markers. Results: Transformation to GEL and the biodegradation were verified. All BaSO{sub 4} markers could be detected in the CBCT. In the OBI and fluoroscopy images, all markers visible in the AP, but only B2(2 mm) could be identified in the lateral images. Changes of BaSO{sub 4} position were not detected in vivo (mice). The volume of the markers decreased by up to 65% and the HU increased by 22%, on average. The mean HU values around the B1, B2, and gold markers were 121.30 [standard deviation (SD): 54.86], 126.31 (SD: 62.13), and 1070.7 (SD: 235.16), respectively. The MC-simulated dose distortion for the BaSO{sub 4} markers was less than that of the commercially used markers. The dose reduction due to the gold marker was largest (15.05%) followed by stainless steel (7.92%) and titanium (6.92%). Dose reduction by B2 (2 mm) was 4.75% and 0.53% in parallel and perpendicular orientations, respectively. Conclusions: BaSO{sub 4} mixed with PBS is a good contrast agent in biodegradable polymer marker because of minimal artifacts in x-ray images and minimal dose reduction in proton therapy. The flexibility of the size is considered to be an advantage of this material over solid type fiducials.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2022

Celeritas

Celeritas is a new Monte Carlo transport code designed for high-performance simulation of high-energy physics detectors.

Johnson, Seth R.↗
DOE OSTI2022

Power distribution estimation method for SMR using ex-core detectors: experimental demonstration by plural control rod patterns at KUCA

The power distribution estimation method based on the ex-core detectors, or PHOEBE, was demonstrated at Kyoto University Critical Assembly, KUCA. Generally, core monitoring systems use in-core neutron detectors. Since inside the core is a harsh environment, the maintainability and reliability of the detectors are deteriorated. On the other hand, the environment outside the core is milder: core monitoring by the ex-core detectors improves the maintainability and reliability especially for small modular reactors and micro-reactors. However, neutron information from the inner region of the core is lost at the ex-core detectors. To recover the information, the authors proposed to utilize the power correlation between the fuel regions. PHOEBE concept was demonstrated at KUCA with distorted power distributions simulated by control rod patterns. The relative power distribution estimated by PHOEBE agreed with that calculated by Monte Carlo simulation code MVP. PHOEBE approximately reproduced the trend of the distorted power distributions calculated by MVP code. By contrast, the case of without power correlation between fuel regions produced significant different power distributions. Therefore, the advantage of considering the power correlation between fuel regions was also demonstrated. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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 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. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

MPACT Fast Neutron Multiplicity System Design Concepts

This report documents work performed by Idaho National Laboratory and the University of Michigan in fiscal year (FY) 2012 to examine design parameters related to the use of fast-neutron multiplicity counting for assaying plutonium for materials protection, accountancy, and control purposes. This project seeks to develop a new type of neutron-measurement-based plutonium assay instrument suited for assaying advanced fuel cycle materials. Some current-concept advanced fuels contain high concentrations of plutonium; some of these concept fuels also contain other fissionable actinides besides plutonium. Because of these attributes the neutron emission rates of these new fuels may be much higher, and more difficult to interpret, than measurements made of plutonium-only materials. Fast neutron multiplicity analysis is one approach for assaying these advanced nuclear fuels. Studies have been performed to assess the conceptual performance capabilities of a fast-neutron multiplicity counter for assaying plutonium. Comparisons have been made to evaluate the potential improvements and benefits of fast-neutron multiplicity analyses versus traditional thermal-neutron counting systems. Fast-neutron instrumentation, using for example an array of liquid scintillators such as EJ-309, have the potential to either a) significantly reduce assay measurement times versus traditional approaches, for comparable measurement precision values, b) significantly improve assay precision values, for measurement durations comparable to current-generation technology, or c) moderating improve both measurement precision and measurement durations versus current-generation technology. Using the MCNPX-PoliMi Monte Carlo simulation code, studies have been performed to assess the doubles-detection efficiency for a variety of counter layouts of cylindrical liquid scintillator detector cells over one, two, and three rows. Ignoring other considerations, the best detector design is the one with the most detecting volume. However, operational limitations guide a) the maximum acceptable size of each detector cell (due to PSD performance and maximum-acceptable per-channel data throughput rates, limited by pulse pile-up and the processing rate of the electronics components of the system) and b) the affordability of a system due to the number of total channels of data to be collected and processed. As a first estimate, it appears that a system comprised of two rows of detectors 5" Ø $\times$ 3" would yield a working prototype system with excellent performance capabilities for assaying Pu-containing items and capable of handling high signal rates likely when measuring items with Pu and other actinides. However, it is still likely that gamma-ray shielding will be needed to reduce the total signal rate in the detectors. As a first step prior to working with these larger-sized detectors, it may be practical to perform scoping studies using small detectors, such as already-on-hand 3" Ø $\times$ 3" detectors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Stochastic Particle Real Time Analyzer (SPARTA) Validation and Verification Suite

This report presents the test cases used to verify, validate and demonstrate the features and capabilities of the first release of the 3D Direct Simulation Monte Carlo (DSMC) code SPARTA (Stochastic Real Time Particle Analyzer). The test cases included in this report exercise the most critical capabilities of the code like the accurate representation of physical phenomena (molecular advection and collisions, energy conservation, etc.) and implementation of numerical methods (grid adaptation, load balancing, etc.). Several test cases of simple flow examples are shown to demonstrate that the code can reproduce phenomena predicted by analytical solutions and theory. A number of additional test cases are presented to illustrate the ability of SPARTA to model flow around complicated shapes. In these cases, the results are compared to other well-established codes or theoretical predictions. This compilation of test cases is not exhaustive, and it is anticipated that more cases will be added in the future.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2014

Stochastic Particle Real Time Analyzer (SPARTA) Validation and Verification Suite

This report presents the test cases used to verify, validate and demonstrate the features and capabilities of the first release of the 3D Direct Simulation Monte Carlo (DSMC) code SPARTA (Stochastic Real Time Particle Analyzer). The test cases included in this report exercise the most critical capabilities of the code like the accurate representation of physical phenomena (molecular advection and collisions, energy conservation, etc.) and implementation of numerical methods (grid adaptation, load balancing, etc.). Several test cases of simple flow examples are shown to demonstrate that the code can reproduce phenomena predicted by analytical solutions and theory. A number of additional test cases are presented to illustrate the ability of SPARTA to model flow around complicated shapes. In these cases, the results are compared to other well-established codes or theoretical predictions. This compilation of test cases is not exhaustive, and it is anticipated that more cases will be added in the future.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

Semi-Analytic Benchmarks for MCNP6

This series of slides shows results for benchmarks for code verifications in MCNP6. For additional related information see report LA-UR--16-28568.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

A Pulsed Sphere Tutorial

Here I attempt to explain what physically happens when we pulse an object with neutrons, specifically what we expect the time dependent behavior of the neutron population to look like. Emphasis is on the time dependent emission of both prompt and delayed neutrons. I also describe how the TART Monte Carlo transport code models this situation; see the appendix for a complete description of the model used by TART. I will also show that, as we expect, MCNP and MERCURY, produce similar results using the same delayed neutron model (again, see the appendix).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Modeling blur in various detector geometries relevant to DARHT

Monte Carlo transport codes have been used to model the detector blur and energy deposition in various detector geometries for MeV radiography which are applicable to DARHT. Segmented scintillator detectors, where low Z scintillators combined with a high-Z metal matrix, are designed such that the resolution increases with increasing metal fraction. The combination of various types of metal intensification screens and storage phosphor imaging plates has also been studied. A storage phosphor coated directly onto a metal intensification screen has superior performance over a commercial plate. Stacks of storage phosphor plates and tantalum intensification screens show an increase in energy deposited and detective quantum efficiency with increasing plate number, at the expense of resolution which degrades with plate number. Excellent agreement between simulation and experimental modulation transfer functions for select detector geometries was found.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

LDRD Report : Analysis of Defect Clustering in Semiconductors using Kinetic Monte Carlo Methods

The transient degradation of semiconductor device performance under irradiation has long been an issue of concern. Neutron irradiation can instigate the formation of quasi-stable defect structures, thereby introducing new energy levels into the bandgap that alter carrier lifetimes and give rise to such phenomena as gain degradation in bipolar junction transistors. Typically, the initial defect formation phase is followed by a recovery phase in which defect-defect or defect-dopant interactions modify the characteristics of the damaged structure. A kinetic Monte Carlo (KMC) code has been developed to model both thermal and carrier injection annealing of initial defect structures in semiconductor materials. Following the development of a set of verification tests, the code is employed to investigate annealing in electron-irradiated, p-type silicon as well as the recovery of base current in silicon transistors bombarded with neutrons at the Los Alamos LANSCE "Blue Room" facility. The results reveal that KMC calculations agree well with experiment once adjustments are made to significant defect parameters within the appropriate uncertainty bounds.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Verification of the prompt radiation shielding design of the ESS Accelerator (Stage 2)

The purpose of this CRADA is to perform additional analyses using MARS Monte-Carlo calculations code for verification of the prompt radiation shielding design of the ESS Accelerator. Stage 2 includes the following activities: 1.Activation assessment of the soil surrounding the wall of the ESS accelerator that maymigrate with the groundwater 2.Shielding assessment of the HVAC exhaust duct The existing ESS accelerator model (with possible minor modifications) shall be used in calculations. This work builds on the analysis originally performed under agreement CRADA FRA-2015-0302.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Shortening the Design and Certification Cycle for Additively Manufactured Materials by Improved Mesoscale Simulations and Validation Experiments: Fiscal Year 2019 Status Report.

This report outlines the fiscal year (FY) 2019 status of an ongoing multi-year effort to develop a general, microstructurally-aware, continuum-level model for representing the dynamic response of material with complex microstructures. This work has focused on accurately representing the response of both conventionally wrought processed and additively manufactured (AM) 304L stainless steel (SS) as a test case. Additive manufacturing, or 3D printing, is an emerging technology capable of enabling shortened design and certification cycles for stockpile components through rapid prototyping. However, there is not an understanding of how the complex and unique microstructures of AM materials affect their mechanical response at high strain rates. To achieve our project goal, an upscaling technique was developed to bridge the gap between the microstructural and continuum scales to represent AM microstructures on a Finite Element (FE) mesh. This process involves the simulations of the additive process using the Sandia developed kinetic Monte Carlo (KMC) code SPPARKS. These SPPARKS microstructures are characterized using clustering algorithms from machine learning and used to populate the quadrature points of a FE mesh. Additionally, a spall kinetic model (SKM) was developed to more accurately represent the dynamic failure of AM materials. Validation experiments were performed using both pulsed power machines and projectile launchers. These experiments have provided equation of state (EOS) and flow strength measurements of both wrought and AM 304L SS to above Mbar pressures. In some experiments, multi-point interferometry was used to quantify the variation is observed material response of the AM 304L SS. Analysis of these experiments is ongoing, but preliminary comparisons of our upscaling technique and SKM to experimental data were performed as a validation exercise. Moving forward, this project will advance and further validate our computational framework, using advanced theory and additional high-fidelity experiments.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Patchy snapshots of nuclear chain reactions

Stochastic fluctuations of the neutron population within a nuclear reactor are typically prevented by operating the core at a sufficiently high power. This regime, where the evolution of the neutron density is essentially deterministic, is key for automatic protection and safety systems to safely detect unwanted power excursions during an accident, and to rapidly initiate a reactor trip procedure in case it is needed. Recent works, supported by numerical simulations, have however reported that, for large reactors, the branching nature of the fission reactions might induce strongly non-Poissonian patterns in the neutron spatial distribution, and that stochastic fluctuations might still persist at reactor powers close to operating conditions (startup phase). An international program conducted by LANL, IRSN and CEA was therefore setup to experimentally detect and characterize such fluctuations and correlations. An experiment took place in 2017 at the Reactor Critical Facility (RCF) of the Rensselaer Polytechnic Institute (USA). In this paper we will report the main findings of this experimental program, supported by stochastic models and by the development of a dedicated high-fidelity Monte Carlo simulation code. We will in particular describe and explain the strong patchiness in neutron power distributions measured at the RCF, as well as a peculiar ‘blinking’ behavior of the core, and discuss the consequences of these findings on nuclear safety.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Transient Thermal Analysis of Calorimeters Used in Characterization of the ACRR Radiation Environments

Silicon calorimeters have been used for active radiation dosimetry in the central cavity of the Annular Core Research Reactor (ACRR) for over a decade. Recently, there has been interest in using other materials for calorimetry to accurately measure the prompt gamma-ray energy deposition in the mixed neutron and gamma-ray environment. The calorimeters used in the ACRR use a thermocouple (TC) to measure the change in temperature of specific materials in the radiation environment. The temperature change is related to the instantaneous dose received by the material in a pulse-transient operation. SOLIDWORKS Simulation and ANSYS Mechanical were used to model the calorimeter and analyze the thermal behavior under pulse-transient conditions. This report compares the results from modeling to experimental results for selected calorimeter materials and radiation environments. These materials include bismuth, tin, zirconium, and silicon. Calorimeters assembled with each material were irradiated in the ACRR central cavity in the free- field, LB44, CdPoly, and PLG radiation environments. The neutronics code Monte-Carlo N- Particle (MCNP) was used to calculate the neutron and gamma-ray response of the calorimeter materials at the experimental locations in the central cavity. Different response tallies were used and found to give different results for the gamma-ray energy deposition. It was determined that performing the neutron/gamma-ray/electron transport in MCNP using the *F8 electron tally gave the overall best agreement with the experimental results. The *F8 tally, however, is much more computationally intensive than the neutron/gamma-ray transport calculations. Also, this report contains parametric analyses that examine the ways to improve the current design of the calorimeters. One finding from the parametric analysis was that the TC should be placed closer to the outer radius of the disks to obtain a measurement closer to the maximum temperature of the disk. Also, the parametric analysis showed that the most dominant mechanism of heat loss in the calorimeters is conduction through the alumina posts. In future designs, the conduction should be minimized to reduce the effect of heat loss on the measurements.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

ADETS User Manual

ADETS (Automated DEpletion Transport System) calculates coupled neutronic/isotopic results for nuclear systems and produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. ADETS is a fully automated tool that links the LANL MCNP Monte Carlo transport code with the SCALE system, specifically ORIGEN-ARP (radioactive decay and burnup code). Additionally, ADETS can compute the dose at various distances of the reactor using a point gamma source, whose characteristics are computed by ORIGEN-ARP. In addition, ADETS is fully coupled with the Uncertanty Quantification, Probabilistic Risk Assessment, Parameter Optimization and Data Analysis code RAVEN (developed at the Idaho National Laboratory as well). RAVEN is released with ADETS but MCNP and SCALE (ORIGEN-ARP) are not included.

42 ENGINEERING↗
DOE OSTI2020

Verification of the CP2020 Library

For many years, “pencil beam” (aka “broomstick”) problems have been used in Monte Carlo neutronic code verification studies . At Los Alamos, they were used to verify the changes associated with the upgrade of NJOY / MCNP to allow continuous angular distributions (instead of discrete angular distributions) from S(α,β) scattering. Recently, “pencil beam” problems have been applied to the verification of charged particle data for MCNP. Another verification test for CP2020 was to compare the continuous energy reaction cross sections generated by ACER with multigroup reaction cross sections generated by GROUPR from the same evaluation file. This method was also used in the verification of the most recent S(α,β) neutron libraries at Los Alamos .

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

YAHFC: A Code Framework to Model Nuclear Reactions and Estimate Correlated Uncertainties

Reaction modeling is a key ingredient in designing experiments and interpreting their results, and is an essential component in the process of evaluating nuclear data and assembling nuclear data libraries used in nuclear technology applications. Typically, experimental data are available only for a handful of reaction channels and theory models are used to fill in the gaps. In addition, theory is often called upon as the arbitrator between discrepant data. Most importantly, theory and modeling are required for an accurate determination of uncertainties in the evaluated data and the correlations between the multiple channels. A fast, accurate, and flexible modeling capability has been developed at LLNL with the code system YAHFC (Yet Another Hauser-Feshbach Code). YAHFC is a Monte Carlo, Hauser-Feshbach code framework, making full use of dynamic memory allocation, derived types, and parallel computing. YAHFC can generate events to simulate experiments and is guiding experiments designed to measure inelastic neutron scattering from actinide targets. YAHFC is also being used to analyze decays from surrogate experiments, thereby enabling the inference of reaction cross sections inaccessible by direct measurement. Finally, by modeling nuclear reactions with constraints from experimental data, YAHFC can deliver complete nuclear data libraries, with evaluated uncertainties, using the modernized Generalized Nuclear Data Structure (GNDS).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗