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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.

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At least 685 records · Page 38
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Validation of statistical uncertainties in subcritical benchmark measurements: Part II – Measured data

Part I of this work presented an uncertainty approach which incorporates the singles and doubles counting rates determined by the Hage-Cifarelli formalism of the Feynman Variance-to-Mean method. This moments-based approach utilizes time correlations between prompt neutrons detections to assess the system multiplication. Additionally, it presented a validation which utilized simulated data generated using a 0-D point-kinetics Monte Carlo code. In Part II, this same method and validation approach is applied to measured data of a subcritical benchmark experiment. These measurements were performed with a 4.5 kg sphere of alpha-phase weapons grade plutonium reflected by copper and/or polyethylene. Finally, the results of 17 configurations are shown including detailed validation for four of the configurations.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2018

Validation of Statistical Uncertainties in Subcritical Benchmark Measurements: Part I - Theory and Simulations

Subcritical neutron noise measurements are often used to determine multiplication of a system with Special Nuclear Material. This work presents an uncertainty approach which incorporates the singles and doubles counting rates determined by the Hage-Cifarelli formalism. This is a moments-based approach which utilizes time correlations between prompt neutrons in order to assess the system multiplication. After the method is described, a validation is presented which utilizes simulated data generated using a 0-D point-kinetics Monte Carlo code.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Utilization of ACE nuclear data file toolkit ACEtk to calculate relative sensitivity coefficients of point-kinetics parameters

Sensitivity and uncertainty methods are quintessential for nuclear criticality safety and experiment design. This type of analysis relies on calculations of sensitivity coefficients; sensitivity coefficients of the effective neutron multiplication factor with respect to some nuclear data are predominantly calculated and used. As a part of the Laboratory Directed Research & Development project EUCLID (Experiments Underpinned by Computational Learning for Improvements in nuclear Data) at Los Alamos National Laboratory, sensitivity coefficients of many radiation detector measurement responses with respect to nuclear data were investigated. Specifically, this paper outlines a method to calculate point-kinetics parameters relative sensitivity coefficients with respect to nuclear data. Point-kinetics parameters such as the prompt neutron decay constant, effective delayed neutron fraction, and neutron generation time are especially important to experimenters and reactor operators designing systems with dynamic neutron populations. This method couples capabilities of the ACE (A Compact ENDF) nuclear data file toolkit, ACEtk, with the ability to load cross sections into the radiation transport code Monte Carlo N-Particle (MCNP). In conclusion, key aspects of optimizing this method for a particular application and sensitivity profiles of the Jezebel criticality experiment are examined and discussed.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2020

Accelerated version of NUBEAM capabilities in DIII-D using neural networks

A neural network model of the effects of neutral beam injection on DIII-D has been developed. The training and testing data used by the model have been generated by the NUBEAM module of TRANSP for experimental discharges from the 2018 DIII-D campaign. Using a principle component analysis to reduce the dimensionality of profile data, the model has been shown to reproduce the results of the Monte Carlo code NUBEAM with a high level of accuracy and an execution time orders of magnitude faster than the execution time of NUBEAM. Furthermore, this makes the neural network model uniquely suited to applications in model-based scenario planning (off-line) and active control (on-line), where a large number of simulation runs are required by the associated optimization tasks that need to be performed before and during the discharge.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2025

Analysis of fast-ion losses measured in MAST-U via infrared thermography and a Fast Ion Loss Detector

Fast-ion losses need to be monitored to avoid damage to plasma facing components. In existing experimental devices, the scintillator-based fast-ion loss detector (FILD) is the most advanced diagnostic for measuring fast-ion losses. However, FILDs provide only local information about the losses. Infrared (IR) thermography can be used as a complementary tool for more global monitoring of the deposition of fast-ion losses on the wall, at the expense of no velocity-space resolution. IR cameras measure the temperature of the plasma facing components. This measurement, determined by a combined effect of the thermal plasma, radiation, neutrons and fast-ion losses, can be decomposed to infer the fast-ion load on the tokamak wall. In this manuscript, a workflow to estimate fast-ion losses via IR thermography is applied to the MAST-U spherical tokamak, using a 1D approximation to extract the experimental heat flux on the FILD front face from IR data. To numerically estimate the different contributions to this total heat flux, the field-line tracing environment SMITER is used to calculate the thermal plasma contribution, the orbit-following Monte-Carlo code ASCOT to estimate the fast-ion losses, and bolometry measurements for the radiation. To validate the workflow, two discharges, L-mode plasmas with low MHD activity, were executed using on and off-axis beams, respectively. The experimentally and numerically estimated heat flux are of the same order of magnitude for the on-axis heated scenario, with a strong dependence of the estimated fast-ion losses contribution on the fit to the kinetic profiles used as input. This is also true for the off-axis heated scenario, where the total numerically estimated heat flux is 2.1 or 1.3 times higher than the maximum experimentally estimated heat flux, depending on the ASCOT input used.

FILD↗
DOE OSTI2026

Preliminary neutronic and thermal analysis of Inertial Fusion Energy systems with a focus on thick liquid wall concepts

This paper provides an assessment of the neutronic behavior of Inertial Fusion Energy systems using the Serpent 2 Monte Carlo code. The study explores different types of reactor concepts, including dry-wall and thick liquid wall systems. In particular, it characterizes the tritium breeding and energy multiplication performance, as well as the effect of long-term irradiation on structural materials. For this purpose, the displacement per atom (DPA) and the gas production rate in structural materials are calculated. A simplified geometry with varying geometrical parameters is used as a representation for the different kinds of devices. Additionally, the case of the HYLIFE-II concept is used to assess modeling inaccuracies arising from certain approximations. A comparison between the simplified geometry and a realistic one is performed, thanks to Serpent’s capability to run transport calculations on unstructured meshes. The study then addresses the impact of these discrepancies on the temperature distribution in the first wall using the GeN-Foam code for conjugate heat transfer analysis. The results highlight the necessity to consider the heterogeneous nature of the liquid wall, as simplified models tend to underestimate radiation damage metrics and heat loads.

Conjugate heat transfer↗
DOE OSTI2012

MCNPX-PoliMi for Nuclear Nonproliferation Applications

In the past few years, efforts to develop new measurement systems to support nuclear nonproliferation and homeland security have increased substantially. Monte Carlo radiation transport is one of the simulation methods of choice for the analysis of data from existing systems and for the design of new measurement systems; it allows for accurate description of geometries, detailed modeling of particle-nucleus interactions, and event-by-event detection analysis. This paper describes the use of the Monte Carlo code MCNPX-PoliMi for nuclear-nonproliferation applications, with particular emphasis on the simulation of spontaneous and neutron-induced nuclear fission. In fact, of all possible neutron-nucleus interactions, neutron-induced fission is the most defining characteristic of special nuclear material (such as U-235 and Pu-239), which is the material of interest in nuclear-nonproliferation applications. The MCNP-PoliMi code was originally released from the Radiation Safety Shielding Center (RSSIC) at Oak Ridge National Laboratory in 2003 [1]; the MCNPX-PoliMi code contains many enhancements and is based on MCNPX ver. 2.7.0. MCNPX-PoliMi ver. 2.0 was released through RSICC in 2012 as a patch to MCNPX ver. 2.7.0 and as an executable [2].

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Evaluation of Am–Li neutron spectra data for active well type neutron multiplicity measurements of uranium

Safeguarding nuclear material is an important and challenging task for the international community. One particular safeguards technique commonly used for uranium assay is active neutron correlation counting. This technique involves irradiating unused uranium with ( α,n) neutrons from an Am-Li source and recording the resultant neutron pulse signal which includes induced fission neutrons. Although this non-destructive technique is widely employed in safeguards applications, the neutron energy spectra from an Am-Li sources is not well known. Several measurements over the past few decades have been made to characterize this spectrum; however, little work has been done comparing the measured spectra of various Am-Li sources to each other. This paper examines fourteen different Am-Li spectra, focusing on how these spectra affect simulated neutron multiplicity results using the code Monte Carlo N-Particle eXtended (MCNPX). Two measurement and simulation campaigns were completed using Active Well Coincidence Counter (AWCC) detectors and uranium standards of varying enrichment. The results of this work indicate that for standard AWCC measurements, the fourteen Am-Li spectra produce similar doubles and triples count rates. Finally, the singles count rates varied by as much as 20% between the different spectra, although they are usually not used in quantitative analysis.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Digital pile-up rejection for plutonium experiments with solution-grown stilbene

A solution-grown stilbene detector was used in several experiments with plutonium samples including plutonium oxide, mixed oxide, and plutonium metal samples. Neutrons from different reactions and plutonium isotopes are accompanied by numerous gamma rays especially by the 59-keV gamma ray of 241Am. Identifying neutrons correctly is important for nuclear nonproliferation applications and makes neutron/gamma discrimination and pile-up rejection necessary. Each experimental dataset is presented with and without pile-up filtering using a previously developed algorithm. The experiments were simulated using MCNPX-PoliMi, a Monte Carlo code designed to accurately model scintillation detector response. Collision output from MCNPX-PoliMi was processed using the specialized MPPost post-processing code to convert neutron energy depositions event-by-event into light pulses. The model was compared to experimental data after pulse-shape discrimination identified waveforms as gamma ray or neutron interactions. We show that the use of the digital pile-up rejection algorithm allows for accurate neutron counting with stilbene to within 2% even when not using lead shielding

Bourne, M. M.↗
DOE OSTI2021

Evaluation of gamma-ray transmission through rectangular collimator slits for application in nuclear fuel spectrometry

Gamma-ray spectrometry is widely applied in several science fields, and in particular in non-destructive gamma scanning and gamma emission tomography of irradiated nuclear fuel. Usually, a collimator is used in the experimental setup, to selectively interrogate a region of interest in the fuel. For the optimization of instrument design, as well as for planning measurement campaigns, predictive models for the transmitted gamma-ray intensity through the collimator are needed. Commonly, Monte Carlo Radiation Transport tools are used for accurate prediction of gamma-ray transport, however, the long computation time requirements when used in low-efficiency experimental setups present challenges. Here, the full-energy peak intensity transmitted through a rectangular collimator slit was examined. A uniform planar surface source emitting isotropically was considered, and the rate of photons reaching an ideal counter plane on the opposite side of the collimator was evaluated by analytical integration. To find a closed-form primitive function, some idealizations were required, and thereby parametric models were obtained for the optical field of view, dependent on slit dimensions (length, height and width) and source-to-collimator distance. For contributions from outside the optical field of view, where a closed-form expression cannot be found, fast numerical integral methods were instead used. The results were validated using the Monte Carlo code MCNP6 and show an agreement within three percent for the numerical method. For the analytical method, deviations up to tens of percent were obtained, which is deemed to still be sufficient for instrument design and measurement planning, where often the order of magnitude of the count rate is not a priori known. The method is planned for use in iterative optimization routines in the design of Gamma Emission Tomography devices, as well as for the prediction of gamma spectra obtained in the planning of fuel inspections. An application of the proposed method was demonstrated in spectrum prediction for a short cooling-time fuel rod test from the Halden reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Measurements and calculations of air activation in the NuMI neutrino production facility at Fermilab with the 120-GeV proton beam on target

Measurements and calculations of the air activation at a high-energy proton accelerator are described. The quantity of radionuclides released outdoors depends on operation scenarios including details of the air exchange inside the facility. To improve the prediction of the air activation levels, the MARS15 Monte Carlo code radionuclide production model was modified to be used for these studies. Measurements were done to benchmark the new model and verify its use in optimization studies for the new DUNE experiment at the Long Baseline Neutrino Facility (LBNF) at Fermilab. The measured production rates for the most important radionuclides – 11 C, 13 N, 15 O and 41 Ar – are in a good agreement with those calculated with the improved MARS15 code.

43 PARTICLE ACCELERATORS↗
DOE OSTI2015

Calculating infinite-medium α-eigenvalue spectra with Monte Carlo using a transition rate matrix method

The time-dependent behavior of the energy spectrum in neutron transport was investigated with a formulation, based on continuous-time Markov processes, for computing α eigenvalues and eigenvectors in an infinite medium. In this study, a research Monte Carlo code called “TORTE” (To Obtain Real Time Eigenvalues) was created and used to estimate elements of a transition rate matrix. TORTE is capable of using both multigroup and continuous-energy nuclear data, and verification was performed. Eigenvalue spectra for infinite homogeneous mixtures were obtained, and an eigenfunction expansion was used to investigate transient behavior of the neutron energy spectrum.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Evaluation of the spatial self-shielding impact for TRISO-based nuclear fuel depletion

Reactor physics analyses of nuclear cores with nuclear fuel concepts containing tristructural isotropic (TRISO) particles, such as pebbles or compact fuel elements, rely on various degrees of simplification to keep these highly heterogeneous problems computationally tractable. One such limitation regards the level of spatial discretization employed during burnup calculations, where traditionally only a limited number of spatial zones are modeled at the full core level and assume that the spectrum is constant within the fuel elements and TRISO particles in this depletion zone. This type of assumption neglects the impact of spatial self-shielding effect within the kernels (microscale level) as well as within the compact or pebbles (mesoscale level). Furthermore, the Monte Carlo code Serpent 2 contains many relevant features for efficiently modeling this type of geometry, including a collision-based domain decomposition intended for very large burnup calculations, which we leveraged for this work to quantify the impact of capturing neutron flux variations occurring at the micro- and mesoscale level on a series of high-temperature gas-cooled reactor fuel element depletion problems. While spatial self-shielding is observed at both scales, with differences from a volume-averaged burnup of ±7% within the kernels and ±2% between TRISO particles within the fuel element, the conjugated effect on nuclide inventories and multiplication factor are negligible, hence confirming that assuming a single average spectrum value may be sufficient for most applications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Laser-Generated Proton Beams for High-Precision Ultra-Fast Crystal Synthesis

We present a method for the synthesis of micro-crystals and micro-structured surfaces using laser-accelerated protons. In this method, a solid surface material having a low melting temperature is irradiated with very-short laser-generated protons, provoking in the ablation process thermodynamic conditions that are between the boiling and the critical point. The intense and very quick proton energy deposition (in the ns range) induces an explosive boiling and produces microcrystals that nucleate in a plasma plume composed by ions and atoms detached from the laser-irradiated surface. The synthesized particles in the plasma plume are then deposited onto a cold neighboring, non-irradiated, solid secondary surface. We experimentally verify the synthesizing methods by depositing low-melting-material microcrystals - such as gold - onto nearby silver surfaces and modeling the proton/matter interaction via a Monte Carlo code, confirming that we are in the above described thermodynamic conditions. Morphological and crystallinity measurements indicate the formation of gold octahedral crystals with dimensions around 1.2 μm, uniformly distributed onto a silver surface with dimensions in the tens of mm 2 . This laser-accelerated particle based synthesis method paves the way for the development of new material synthesis using ultrashort laser-accelerated particle beams.

Science & Technology - Other Topics↗
DOE OSTI2024

Preliminary analysis of TREAT free-field experiments using openmc

This work analyses activation calculations for dosimetry materials during a steady-state irradiation in the Transient Reactor Test (TREAT) reactor core. Hence, we developed a workflow based on the Monte Carlo code OpenMC alongside a custom depletion solver. The irradiation-induced activity as a function of time is computed, and several sensitivity studies are performed to evaluate uncertainty. This study has shown activity computations are sensitive to flux amplitude, irradiation time, atoms quantity and microscopic cross sections. Stochastic uncertainties have been propagated to evaluate the activity uncertainty for each dosimetry material. Most uncertainties are below our target of 3%, which demonstrates OpenMC as a powerful predictive and analysis tool. The precise results obtained through this newly developed computation scheme will be used in future experiments to characterize quantities of interest when operating the TREAT reactor in new configurations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Preservation of kinetics parameters generated by Monte Carlo calculations in two-step deterministic calculations

The generation of accurate kinetic parameters such as mean generation time Λ and effective delayed neutron fraction β eff via Monte Carlo codes is established. Employing these in downstream deterministic codes warrants another step to ensure no additional error is introduced by the low-order transport operator when computing forward and adjoint fluxes for bilinear weighting of these parameters. Another complexity stems from applying superhomogenization (SPH) equivalence in non-fundamental mode approximations, where reference and low-order calculations rely on a 3D full core model. In these cases, SPH factors can optionally be computed for only part of the geometry while preserving reaction rates and K-effective, but the impact of such approximations on kinetics parameters has not been thoroughly studied. This paper aims at studying the preservation of bilinearly-weighted quantities in the Serpent–Griffin calculation procedure. Diffusion and transport evaluations of IPEN/MB-01, Godiva, and Flattop were carried out with the Griffin reactor physics code, testing available modeling options using Serpent-generated multigroup cross sections and equivalence data. Verifying Griffin against Serpent indicates sensitivities to multigroup energy grid selection and regional application of SPH equivalence, introducing significant errors; these were demonstrated to be reduced through the use of a transport method together with a finer energy grid.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗