Search NASA⌕ Search

SEARCH · Search NASA

Results for “MCNP Model”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

As-run physics analysis for the EPRI-2 experiment for cycle 153B

The purpose of this ECAR is to document the as-run physics analysis and source term in Curies and decay heat rate (Watts) for the shipment of the EPRI-2 experiment located in the Center Flux Trap (CFT). EPRI-2 was irradiated during cycle 153B. The experiment irradiation ended on April 13, 2013. The as-run heat rates and flux/fluence results were calculated using the MCNP ATR full core model. The heat rates and flux are calculated based on an average center lobe power of 30.8 MW. The fluence is based on the operating time of 13.45 days. ORIGEN2 Version 2.2 is used to determine the decay heat rate, gamma spectrum, and source term for the EPRI-2 experiment. The source term analysis was performed as requested by the project staff to demonstrate compliance with shipping requirements in the GE-2000 cask following irradiation as well as provide an estimated source term to support Post-Irradiation Examination as needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Optimization of a neutron detector design using adjoint transport simulation

A synthetic aperture approach has been developed and investigated for Special Nuclear Materials (SNM) detection in vehicles passing a checkpoint at highway speeds. SNM is postulated to be stored in a moving vehicle and detector assemblies are placed on the road-side or in chambers embedded below the road surface. Neutron and gamma spectral awareness is important for the detector assembly design besides high efficiencies, so that different SNMs can be detected and identified with various possible shielding settings. The detector assembly design is composed of a CsI gamma-ray detector block and five neutron detector blocks, with peak efficiencies targeting different energy ranges determined by adjoint simulations. In this study, formulations are derived using adjoint transport simulations to estimate detector efficiencies. The formulations is applied to investigate several neutron detector designs for Block IV, which has its peak efficiency in the thermal range, and Block V, designed to maximize the total neutron counts over the entire energy spectrum. Other Blocks detect different neutron energies. All five neutron detector blocks and the gamma-ray block are assembled in both MCNP and deterministic simulation models, with detector responses calculated to validate the fully assembled design using a 30-group library. The simulation results show that the 30-group library, collapsed from an 80-group library using an adjoint-weighting approach with the YGROUP code, significantly reduced the computational cost while maintaining accuracy. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

UPF Data Book: A Resource for Standard Data, Dimensions, Equations, and Practices used at UPF for Nuclear Criticality Safety Analyses

A new Uranium Processing Facility (UPF) is being designed at Y-12 in Oak Ridge to replace the existing facilities. As part of this design, new and existing processes need to be evaluated using Criticality Safety Process Studies (CSPS) to ensure they remain subcritical for both normal and credible abnormal conditions. This requires a great many sets of calculations using both MCNP and SCALE to model the various configurations, concentrations, and processes being designed. At UPF these calculations are contained within Design Analysis and Calculation (DAC) documents. Without a standard guide different nuclear criticality safety (NCS) engineers were modeling the same system using parameters familiar to them from past experience. Often this resulted in significantly different assumptions on things such as: materials, uranium enrichment, reflection conditions, and maximum uranium concentrations for calculations in DACs supporting essentially the same process but in a different CSPS. It became apparent that, for DACs to be useful for more than the CSPS they were written to support, a common set of guidelines, for use by all NCS engineers, was needed. A Data Book DAC is now used at UPF to help standardize the common UPF system parameters used for all calculations. This makes it possible to use existing calculations from one CSPS in another CSPS having similar processes. It also simplifies the requirements imposed by NCS on things like pipe diameters, tank size and spacing, and component placement on the design and process disciplines. Finally, for a given parameter, once it is determined the parameter is valid for a UPF process it becomes valid for all similar processes without the need for further calculational analyses. (authors)

42 ENGINEERING↗
DOE OSTI2024

Embedding Neural Thermal Scattering (NeTS) Modules in SERPENT for Higher Fidelity Advanced Reactor Analysis

When a neutron born in fission thermalizes to the order of $k$ $B$ $T$, it’s de-Broglie wavelength and energy approach the order of inter-atomic spacing and elementary lattice oscillations, respectively. $S$($a,β,t$) or the scattering law, uuantify these temperature-dependent crystallographic contributions to total cross section (or reaction rate). In a Monte Carlo analysis, cumulative distribution functions (CDFs) of $S$($a,β,t$) are loaded to memory from “A Compact ENDF” (ACE) files for stochastically selecting thermal scattered neutron trajectories. In this work, novel neural thermal scattering (NeTS) modules for $S$($a,β,t$) CDFs are designed, trained, serialized and embedded within SERPENT using Python’s limited C-API for on-the-fly deployment of crystalline graphite $S$($a,β,t$) sampling. Torchscript tracing and Numba just-in-time (JIT) compilation streamline neural inference on NVIDIA GPUs with CUDA libraries. Demonstrations of bare sphere thermalization of fast and thermal sources show excellent agreement between embedded NeTS in SERPENT and MCNP. With an explicit model of the reactor, NeTS can predict on-the-fly changes in TREAT neutron spectra as a function of local temperature, which can serve to improve transient and accident predictions in a multiphysics analysis framework. This framework can be further extended to account on-the-fly for changes in local graphitic microstructure to scattering cross sections, and outlines a novel coupling of modern machine learning with state-of-the-art reactor physics methods.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Simulation of gallium nitride vertical fin-shaped field effect transistor for use as thermal neutron detector

Through the use of a radiation detection system simulation framework, a gallium nitride vertical fin-shaped field effect transistor (FinFET) was studied for output response when utilized as a thermal neutron detector. The FinFET was assumed to had been backfilled with boron carbide, reactive to thermal neutrons. The GaN FinFET was modeled with radiation transport from MCNP, and the electronic transport from COMSOL Multiphysics. Fabricated FinFET devices (not neutron reactive) were tested to aid in the tuning of the COMSOL FinFET model. Through time-depenent studies, the drain current response pulse to simulated ionization due to single events lead to building of a database of device responses to radiation. By integrating the current pulses over time, the induced charge was calculated. Using the results of the radiation transport PTRAC file in combination with the induced charge database, an integrated charge spectrum was calculated.

Davidson, Bryce L.↗
DOE OSTI2019

Big Ten MCNP6 Unstructured Mesh Benchmark

The purpose of this report is to document a benchmarking effort for the MCNP6 Unstructured Mesh (UM) geometry description method applied to a large and fairly complex critical assembly, Big Ten. A secondary goal of this effort is to provide a set of best practices for modeling a critical assembly with the MCNP UM method. Specification of geometry using UM requires creating a mesh from a solid geometry model and there are some important considerations when using this method for criticality problems. These considerations will be discussed along with suggested practices for correctly modeling these types of problems.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2026

Rapid Gamma Simulations of TRISO Fuel Elements

As energy demand rises, nuclear energy, particularly from reactors that use tristructural isotropic (TRISO) fuels, has gained attention due to the fuel’s enhanced resistance to radiation damage and high temperatures. This report investigates the modeling capabilities of the Gamma Detector Response and Analysis Software (GADRAS) for TRISO fuels, focusing on the gamma signatures of TRISO particles, which have not been extensively explored. Using the Monte Carlo N-Particle (MCNP) code as a benchmark, we developed both homogeneous and heterogeneous models of TRISO pebbles to analyze gamma spectra. Our findings reveal that the homogeneous and heterogeneous models produced different gamma signatures. Additionally, the GADRAS heterogeneous model significantly reduces computation times compared to MCNP, enabling effective modeling of gamma signatures for safeguards applications. This advancement is essential for the International Atomic Energy Agency (IAEA) in detecting anomalies and potential smuggling attempts in TRISO reactor fuel elements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

PWR commercial reactor criticals as validation for spent fuel transport cask burnup credit - 14489

The burnup credit approach for a spent fuel transportation cask (SFTC) with a high density basket for 32 PWR fuel assemblies required credit not only for actinides but also for fission products, otherwise the required minimum burnups of fuel assemblies would have been too high to be practical. At the time of the development of the burnup credit methodology for this condition, no established approach existed to quantitatively validate credit for fission products. A validation approach was therefore developed specifically for this application, based on Commercial Reactor Criticals (CRCs). In this approach, hot critical state points of PWR plants were evaluated using the same codes that were used for the burnup credit evaluations for the transportation cask, namely CASMO-4 and MCNP4. A total of 36 of those state-points were analyzed, from 5 plants with two different assembly types. For each axial section of each fuel assembly, a separate CASMO-4 depletion analysis was performed, which required a total of about 4000 individual CASMO runs. The spent fuel compositions were then extracted from the CASMO runs and transferred to MCNP, where each core was modeled, considering the existing 1/8 symmetry. The core operating conditions were taken from work previously performed and documented for the Yucca Mountain project. The results of the MCNP calculations were then analyzed to establish bias and bias uncertainty values. Additional studies were performed to compare the in-core condition for the fuel with the condition that exists for the fuel in the spent fuel transport cask. The paper presents and discusses the results of the calculations, and the analysis of those results to establish bias and bias uncertainty values. The paper also presents results of the studies comparing in-core with transport cask conditions, and discusses the resulting limitations of this approach. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

The Preliminary Validation and Verification of GSM

The Generalized Spallation Model, or GSM, simulates hadron-nucleus and nucleus-nucleus collisions at intermediate to high incident energies. It is recommended that use of GSM be limited to interactions for particles and nuclei with kinetic energies greater than ~ 100 MeV, up to ~ 1 T eV/A. GSM was developed to investigate the deprecation of the Los Alamos Quark-Gluon String Model (LAQGSM), as implemented in the event generator LAQGSM03.03. The deprecation of LAQGSM will allow for a simplification of the high energy physics models utilized within the Monte Carlo N-Particle® (MCNP®) transport code1 . MCNP applies simulated data from the Cascade Exciton Model (CEM), as implemented in the CEM03.03 event generator, and the LAQGSM event generator for most high energy reactions, by default, for its simulations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Predictive Modeling of Terrestrial Radiation Exposure from Geologic Materials

Aerial gamma ray surveys are important for those working in nuclear security and industry for determining locations of both anthropogenic radiological sources and natural occurrences of radionuclides. During an aerial gamma ray survey, a low flying aircraft, such as a helicopter, flies in a linear pattern across the survey area while measuring the gamma emissions with a sodium iodide (NaI) detector. Currently, if a gamma ray survey is being flown in an area, the only way to correct for geologic sources of gamma rays is to have flown the area previously. This is prohibitively expensive and would require complete national coverage. This project’s goal is to model the geologic contribution to radiological backgrounds using published geochemical data, GIS software, remote sensing, calculations, and modeling software. K, U and Th are the three major gamma emitters in geologic material. U and Th are assumed to be in secular equilibrium with their daughter isotopes. If K, U, and Th abundance values are known for a given geologic unit the expected gamma ray exposure rate can be calculated using the Grasty equation or by modeling software. Monte Carlo N-Particle Transport software (MCNP), developed by Los Alamos National Laboratory, is modeling software designed to simulate particles and their interactions with matter. Using this software, models have been created that represent various lithologies. These simulations randomly generate gamma ray photons at energy levels expected from natural radiologic sources. The photons take a random path through the simulated geologic media and deposit their energy at the end of their track. A series of nested spheres have been created and filled with simulated atmosphere to record energy deposition. Energies deposited are binned in the same manner as the NaI detectors used during an aerial survey. These models are used in place of the simplistic Grasty equation as they take into account absorption properties of the lithology which the simplistic equation ignores.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Photon detector response function methodology using MCNP and shift hybrid radiation transport code for wide-area contamination assay applications

Here, radiation transport modeling using the Monte Carlo N-Particle (MCNP) radiation transport code and Monte Carlo code, Shift, were employed to model detector responses for a variety of wide-area photon contamination scenarios. In this study, 2" × 2" and 3" × 3" cylindrical NaI(Tl) scintillation detector configurations at source detector-distances of 0.5 cm, 1 cm, 2.54 cm, 10 cm, and 30 cm were modeled. Media of soil, concrete, and steel were evaluated for contamination depths ranging from surface to a depth of an infinite thickness in each medium for photon energies ranging from 20 keV to 3 MeV, which correspond to the energies that current detectors can discern. Monoenergetic photon surface contamination detector responses for each of the media, source–detector distances, and detectors were estimated using MCNP v6.2. Shift was harnessed for improved variance reduction of particle transport in highly attenuating media to obtain average cell fluxes in the two MCNP NaI(Tl) scintillation detector configurations. Average cell flux values in Shift were coupled with detector responses from MCNP to convert average cell flux in a void to energy distribution of pulses in the NaI(Tl) scintillation detector crystal of interest. An optimized detector response function methodology was developed by coupling the MCNP radiation transport method with the Consistent Adjoint Driven Importance Sampling (CADIS) hybrid radiation transport method built into Shift to significantly decrease the runtime of thousands of MCNP pulse height simulations. The methodology may be utilized to quickly and accurately facilitate the assessment of a broad range of wide-area environmental contamination assay and decommissioning cleanup applications.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2026

On the Statistical Uncertainty of Monte Carlo-Calculated Scattering Sensitivities

Sensitivity coefficients calculated with Monte Carlo codes are widely used for nuclear data uncertainty quantification in the modeling and simulation of complex 3D reactor systems. This study systematically compares sensitivity coefficients and associated statistical uncertainties for the multiplication factor and fuel temperature reactivity across multiple Monte Carlo codes (SCALE/KENO, SCALE/Shift, MCNP, and Serpent) using simple models representing light-water reactors and advanced reactor concepts. For multiplication factor sensitivities, statistical uncertainties are generally acceptable, although scattering sensitivities show significantly larger statistical uncertainties than, for example, fission and capture reactions. Fuel temperature reactivity sensitivities show significantly larger statistical uncertainties across all reactions. Elastic scattering sensitivities are the most problematic: all Monte Carlo codes fail to resolve energy-dependent coefficients, and they produce dramatically different energy-collapsed values. Critically, the use of these sensitivity coefficients in nuclear data uncertainty propagation leads to reduced statistical uncertainties in individual uncertainty contributions. This can lead to the masking of unusable sensitivity coefficients and producing misleading uncertainty results. The findings of this study show that new or enhanced methods are needed to improve Monte Carlo elastic scattering sensitivity calculations. Additionally, this study shows the relevance of verifying sensitivity coefficients through direct perturbation calculations for individual nuclide reactions, instead of only for total cross sections as commonly done.

Bostelmann, Rike [ORNL] (ORCID:0000000165968088)↗
DOE OSTI2016

Examining Computational Assumptions For Godiva IV

Over the course of summer 2016, the effects of several computational modeling assumptions with respect to the Godiva IV reactor were examined. The majority of these assumptions pertained to modeling errors existing in the control rods and burst rod. The Monte Carlo neutron transport code, MCNP, was used to investigate these modeling changes, primarily by comparing them to that of the original input deck specifications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Monte Carlo simulation of a Bonner sphere spectrometer for application to the determination of neutron field in the Experimental Advanced Superconducting Tokamak experimental hall

To assess the neutron energy spectra and the neutron dose for different positions around the Experimental Advanced Superconducting Tokamak (EAST) device, a Bonner Sphere Spectrometer (BSS) was developed at Peking University, with totally nine polyethylene spheres and a SP9 {sup 3}He counter. The response functions of the BSS were calculated by the Monte Carlo codes MCNP and GEANT4 with dedicated models, and good agreement was found between these two codes. A feasibility study was carried out with a simulated neutron energy spectrum around EAST, and the simulated “experimental” result of each sphere was obtained by calculating the response with MCNP, which used the simulated neutron energy spectrum as the input spectrum. With the deconvolution of the “experimental” measurement, the neutron energy spectrum was retrieved and compared with the preset one. Good consistence was found which offers confidence for the application of the BSS system for dose and spectrum measurements around a fusion device.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRS1997

Assay of the Martian Regolith with Neutrons

The purpose of the research is to combine experiments and Monte Carlo transport of neutrons through volume of soil in an attempt to model neutron leakage from planetary surfaces. Emphasis is given to the change of neutron spectra as a function of water content and location. During the first stage of effort, two experiments were conducted in which leakage of neutrons from a Pu-Be source through about 30 g/cm(exp 2) of soil were measured with several counters. A Monte Carlo code, MCNP, has been used to model many of the 100 individual runs of the experiment. Hydrogen is the element that has the most dramatic effect on the neutron spectrum and its effect on the neutron spectrum is almost the same whether it is in the form of water or polyethylene. In order to simulate various water configurations, sheets of polyethylene have been used between layers of soil as well as water in several concentrations up to 18%. Comparison of experimental results to theoretical predictions made with the MCNP code were disappointing for low concentrations of water. We have made extensive calculations to see if room return could be the cause of the discrepancies. Water concentrations of the 'dry' soil were measured by two different laboratories and differed only by 0.5%. We have made calculations to optimize the next experiment and are investigating other methods of determining the water content of 'dry' soil.

Drake, Darrell M.↗
DOE OSTI2012

Modeling filters for formation of mono-energetic neutron beams in the research reactor IRT MEPhI and optimization of radiation shielding for liquid-xenon detector

The paper considers formation of mono-energetic neutron beams at the entrance of experimental channels in research reactors for various applications. The problem includes the following steps: 1. Full-scale mathematical model of the research IRT MEPhI was developed for numerical evaluations of neutron spectra and neutron spatial distribution in the area of experimental channels. 2. Modeling of filters in the channel to shift neutron spectrum towards the required mono-energetic line was performed. 3. Some characteristics of neutron beams at the entrance of detector were evaluated. The filter materials were selected. The calculations were carried out with application of the computer code based on the high-precision Monte-Carlo code MCNP. As a result, mathematical model was created for the filter which is able to form mono-energetic (24 keV) neutron beam. The study was carried out within the frames of the research project on development of Russian emission detector with liquid noble gas to observe rare processes of neutrino scattering and particles of hypothetical dark matter in atomic nuclei. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

An unstructured mesh based neutronics optimization workflow

We have developed a fully automated workflow to optimize the neutronics performance of the Second Target Station (STS) at the Oak Ridge National Laboratory’s Spallation Neutron Source. The optimization workflow starts with the parametrized solid CAD engineering models and converts them into the unstructured mesh (UM) models for the neutronics calculations with MCNP6.2. Calculations are executed and their results are loaded into the Dakota optimization toolkit. Dakota analyzes the results and proposes new geometry parameters for the next design iteration. The cycle repeats until the optimal parameters are found. The automated CAD to MCNP conversion, the use of high-fidelity UM models, and the use of modern optimizer are the key elements that advance the entire optimization workflow in comparison with the original workflow. The original workflow was based on a simplified constructive solid geometry (CSG) modeling with MCNPX, mcnp_pstudy tool, and an in-house optimizer. Herein to demonstrate the new workflow, we present a case of neutronics optimization of the moderator–reflector assembly (MRA). Apart from the MRA, the workflow can optimize other major STS components, such as the spallation target, neutron beamlines, radiation shielding, and various accelerator components. Importantly, the new workflow opens the door to the advanced multi-physics multi-parameter optimization and has the potential for use in other nuclear physics and accelerator applications.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Verification and validation of atmospheric transport models with the UF Training Reactor

The University of Florida Training Reactor (UFTR) is a 100-kW ARGONAUT type reactor with a unique design such that a large volume of air passes through the core, producing and releasing significant quantities of 41 Ar. Once released, the radioargon is dispersed throughout the atmosphere and decays by a 1294 keV gamma ray emission. This quantified, validated, and controlled 41 Ar plume can be utilized to verify and validate atmospheric transport models (ATMs). This is accomplished by coupling an ATM with MCNP and comparing the results to experimental measurements. For this study, an in-house developed Gaussian dispersion model and the EPA approved Gaussian model AERMOD were coupled with MCNP to simulate the results of a six hour measurement taken on May 8th, 2018. For the measurements and simulations, a grid of seven NaI detectors were placed in four different rooftop locations near the UFTR plume stack. With easterly winds dominating during the morning and early afternoon, both the measurement results and simulations displayed a higher 41 Ar photon count rate on the detectors west of the plume stack throughout the day. Overall, the AERMOD simulations produced better results when compared to the in-house model, particularly for the detectors placed east and west of the plume stack. While some of the AERMOD data showed a strong correlation with the measurements, the results remain inconsistent. Here, due to the varying accuracy of the simulation results, it is theorized that the relationship between measurement results and simulations will improve with the use of localized weather information or the utilization of non-Gaussian dispersion models.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗