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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 325 records · Page 18
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

A holistic approach to the optimization of neutron beam transport at the LANSCE facility

An advanced neutron beam collimation system has been developed as part of an effort to optimize neutron beam transport at the Weapons Neutron Research (WNR) Facility, located within the Los Alamos Neutron Science Center (LANSCE). The goal of this work was to develop and demonstrate techniques to tailor neutron delivery in order to provide maximum available flux on sample with a specific beam profile, while simultaneously reducing unwanted background. A holistic approach was taken, upgrading the facility spallation target, facility metrology infrastructure, and flight path shutter insert to support the implementation of this advanced collimation system. Here, modern instruments and software were employed to conduct facility surveys, characterization of as-built geometry of critical components, 3D ray tracing and neutron transport calculations. Beam images and flux measurements were acquired after installation to evaluate the performance of the collimation system and to demonstrate consistent agreement with MCNP simulations which showed an 87% increase of flux on sample while providing suppression of background neutron impingement on the sample frame by 10 3 .

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Lithium indium diselenide — An advanced material for neutron detection

Here, this paper describes the synthesis, crystal growth, detector fabrication, radiation hardening studies, MCNP modeling, and characterization of lithium indium diselenide or LiInSe 2 . This newly-developed room temperature thermal neutron detector has semiconducting and scintillating properties and it is suitable for neutron detection application. LiInSe 2 was synthesized starting from elemental Li, In, Se in two steps due to high reactivity of Li. A single crystal of LiInSe 2 was grown using the Vertical Bridgman method. The room temperature band gap was found to be 2.8 eV using optical absorption measurements. Bulk resistivity was measured at ∼5 × 10 11 Ωcm. Photoconductivity measurements of LiInSe 2 wafers identified a peak in the photocurrent around 445 nm. Nuclear radiation detectors were fabricated from single crystal wafer and the responses to alpha particles at various biases were measured. The mobility-lifetime product was estimated. Gamma irradiation studies were performed with calculated absorbed doses ranging from 0.2126 to 21,262 Gy. The characterization of the two wafers for their scintillator performance was conducted after each irradiation. The gamma irradiation produced a reduction of the light yield that translated to a lower channel number for the centroid of alpha detection spectra. It also showed a considerable reduction of the decay time after the first irradiation. These are the first studies on gamma radiation hardening with this material.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Experimental demonstration and modeling of a robotic neutron detector with spectral and directional sensitivity for treaty verification

International safeguards and arms control agreements often require labor-intensive, intrusive onsite inspections to perform verification tasks. The ability to localize a neutron source and/or characterize a neutron field may be imperative for identifying anomalies. We are interested in the role of autonomous mobile robots, which, if designed properly, may be more effective and efficient and less intrusive than their human counterparts. Toward developing such a capability, we, for this study, have constructed the N-SpecDir Bot, comprised of three boron-coated straw detectors azimuthally-distributed within a cylinder of high-density polyethylene, which is mounted on an omni-directional robotic platform. Our N-SpecDir Bot is specifically designed to provide spectral and directional sensitivity, in addition to gross counts, by utilizing the signals from the three detectors. The detection system has been extensively characterized by MCNP modeling, which has been benchmarked to experiments conducted at the Princeton Plasma Physics Laboratory. We demonstrate the spectral and directional sensitivity experimentally and in simulation, and provide a simple yet robust method for directional measurements.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2024

Spherical time-encoded radiation imaging simulations

Radiation source localization is important for nuclear nonproliferation and can be obtained using time-encoded imaging systems with unsegmented detectors. A scintillation crystal can be used with a moving coded-aperture mask to vary the detected count rate produced from radiation sources in the far field. The modulation of observed counts over time can be used to reconstruct an image with the known coded-aperture mask pattern. Current time-encoded imaging systems incorporate cylindrical coded-aperture masks and have limits to their fully coded imaging field-of-view. This work focuses on expanding the field-of-view to 4π by using a novel spherical coded-aperture mask. A regular icosahedron is used to approximate a spherical mask. This icosahedron consists of 20 equilateral triangles; the faces of which are each subdivided into four equilateral triangle-shaped voxels which are then projected onto a spherical surface, creating an 80-voxel coded-aperture mask. Furthermore, these polygonal voxels can be made from high-Z materials for gamma-ray modulation and/or low-Z materials for neutron modulation. In this work, we present Monte Carlo N-Particle (MCNP) simulations and simple models programmed in Mathematica to explore image reconstruction capabilities of this 80-voxel coded-aperture mask.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Applied radiation tags for warhead dismantlement transparency

Here, this study explores the feasibility of applying radiation tags as a Chain of Custody tool to confirm the dismantlement of individual warheads by investigating time dependent emissions from special nuclear material. Precedents of photon and neutron sources that have been used to irradiate warhead components are presented. Measurements of irradiated Highly Enriched Uranium (HEU) are used to benchmark simulations with the CINDER90 library for MCNP. Delayed gamma emissions are simulated as a function of time since irradiation of moderated HEU and Weapons Grade Plutonium with photon and neutron sources. Models of commercial 9 MV and 15 MV linacs were found to induce ~10 9 fissions/g and ~10 11 fissions/g, respectively, creating 3 to 4 orders of magnitude more delayed gammas than neutron sources, modeled at 10 10 n/s. However, a DT neutron source was found to induce sufficient fissions, ~10 7 fissions/g, to create quantifiable signatures relative to intrinsic emissions. Applied radiation tagging is determined to be technically feasible, but its operational context and practical implementation requires significant development.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2025

Characterizing the Experiment for Calibration with Uranium (Excalibur) neutron source for use in warhead verification

Neutron sources can play a variety of roles in warhead verification. For transmission radiography, a source of directed high energy neutrons is required, while for applications to detect fissile isotopes, sub-MeV neutrons are preferred. The Excalibur (Experiment for Calibration with Uranium) neutron source has been built and used in a variety of verification-related experiments. Excalibur is based on a commercial deuterium-tritium neutron generator specified and measured to be capable of producing 14 MeV neutrons at rates of up to 8.2 × 10 8 neutrons/s. Here, the generator is enclosed in a carbon-steel 32" diameter, 23.62" high carbon-steel cylinder that moderates the mean neutron energy to under 500 keV. This, in turn, is encased in 5%-borated polyethylene such that the entire assembly is a 48" x 48" box that is 30" tall. For radiographic applications, a narrow, tapered channel in the steel and polyethylene allows 14 MeV neutrons to stream directly from the generator to a test object. Its collimating capability is demonstrated by measuring the neutron flux profile. In the moderated mode of operation, the generator is fully enclosed in the steel, but a large section of the polyethylene is removed, providing a flux of sub-MeV neutrons from a wide range of angles. Neutron angular and spectral measurements using both a nested neutron spectrometer and a commercial liquid scintillator coupled with a 3 He detector show the expected softer neutron spectrum in moderated mode in good agreement with MCNP6 calculations. The gamma-ray spectrum from Excalibur is also in good agreement with MCNP modeling. Based on these findings, the future application of Excalibur in its two configurations is discussed.

Fissile material detection↗
DOE OSTI2025

Optimizing a detection system for fissile material in nuclear disarmament verification

In arms control treaties, verification plays a crucial role in detecting non-compliance, deterring future violations, and building trust between state parties. Neutron interrogation that induces fission reactions in fissile isotopes and measures the resulting fission neutrons, could be employed for this purpose. This study aims to develop a system which can determine the presence of fissile material while intrinsically protecting information. In this paper, we focus on optimizing the system for discriminating between an enriched uranium block from a depleted uranium (DU) block. The system was built and we report on benchmark measurements with DU and 16% enriched uranium blocks. Furthermore, the Excalibur (Experiment for Calibration with Uranium) neutron source, a neutron spectrometer (redBubble Technology Industries (BTI) N-Probe), and superheated droplet detectors were used for these measurements. MCNP simulations provided insights into detector responses to fissile materials with varying isotopic compositions, confirming that the system functioned as designed.

Active neutron interrogation↗
DOE OSTI2020

Methodology for inferring reactor core power distribution from an optical fiber based gamma thermometer array

Here, a 4-step data analytic methodology has been devised for the purpose of inferring the distribution of power in a reactor core, based on the response of an array of optical fiber based gamma thermometers (OFBGTs). This data analytic methodology is crucial for the development of a system of OFBGTs for the purpose of calibrating local power range monitors in boiling water reactors. Such a system would be an improvement to the present calibration system in boiling water reactors, in terms of safety, efficiency, and permanence. The first step of this methodology is to establish an energy balance method. In this method, one uses MCNP to determine response functions, which allow one to convert from gamma thermometer response to power. The gamma thermometers and the reactor core are segmented, such that each gamma thermometer segment provides an estimate of power for each reactor core segment. The estimates of power for the reactor core segments (hereafter referred to as fuel assembly segments) are calculated based on the measured response of the gamma thermometers. The second step of the methodology is to employ a weighting scheme to combine the estimates of the power of the various fuel assembly segments, based on the response of the various OFGBT segments and the incremental dose rates of the various fuel assembly segments to the OFBGT segments. The third step is to iteratively calculate the estimates of the power of the fuel assembly segments, until a convergence criterion is met, which indicates that the calculation has converged. The fourth and final step of the methodology is to estimate the uncertainty of the power for each fuel assembly segment. The mathematical basis for this step is not the focus of this paper. We have used a 3-D homogeneous reactor model to demonstrate the data analytic methodology; and have found that the data analytic methodology operates as intended.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

An approach towards plastic scintillators from thermally activated delayed fluorescent dyes and cross-linkable bismuth compounds

Cross-linked polyvinyltoluene (PVT) based plastic scintillators loaded with dyes exhibiting thermally activated delayed fluorescence (TADF) and various amounts of (caproate)di-(methacrylate) bismuth (CMB) were developed. The co-polymerization of vinyl toluene monomer loaded with different weight ratios of CMB and 1 wt% of TADF dye yielded highly transparent cross-linked plastic scintillators with intense green photoluminescence. The thermal and mechanical properties of the scintillators were investigated using thermogravimetric analysis (TGA), differential scanning calorimetry (DSC) and micro indentation techniques. Here, their performance for radiation detection was assessed by performing gamma spectroscopy using a Cs-137 source and benchmarked to commercial plastic scintillators. The loading of CMB was guided by Monte Carlo N-Particle (MCNP) simulations. The results of this study show that the cross-linking approach and the use of TADF dyes yield scintillators with good mechanical properties with a uniaxial yield strength up to ~66 MPa for compositions loaded with 40 wt% of CMB and high optical quality. However, the light yield in gamma spectroscopy experiments is found to be reduced with the addition of CMB, suggesting a trade-off between mechanical yield strength and light yield in gamma spectroscopy.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

An Improved Energy Deposition Model in MPACT and Explicit Heat Generation Coupling with CTF

The default energy deposition model in the CASL neutronics code MPACT assumes all fission energy is deposited locally in fuel rods. Furthermore, equilibrium delayed energy release is assumed for both steady-state and transient calculations. These approximations limit the accurate representation of the heat generation distribution in space and its variations over time, which are essential for power distribution and thermal-hydraulic coupling calculations. In this paper, an improved energy deposition model is presented in both the spatial and time domains. Spatially, the energy deposition through fission, neutron capture, and slowing-down reactions are explicitly modeled to account for the heat generation from all regions of a reactor core, and a gamma smearing scheme is developed that utilizes the gamma sources from neutron fission and capture. In the time domain, the delayed energy release is modeled by solving an additional equation of delayed heat emitters, similar to the equation of delayed neutron precursors.To allow the explicit heat generation coupling, the interfaces between MPACT and CTF were updated to transfer separate heat sources for different material regions (fuel, clad, moderator and guide tube). The results show that the distributions of the energy deposition between MPACT and MCNP agree very well for various 2-D assembly and quarter-core problems without TH feedback. The MPACT/CTF coupled calculation for the hot full power quarter-core case exhibited a reduced peak pin power by 2.3% and a reduced peak fuel centerline temperature by 17 K when using the explicit energy deposition and heat transfer. The new model also shows a maximum 100 pcm keff effect on assembly depletion problems and an increased overall energy release by 7% in a PWR reactivity-initiated accident (RIA) problem.

Liu, Yuxuan↗
DOE OSTI2023

The Covariance of PFNS Results from the Chi-Nu Experiment

The prompt fission neutron spectrum (PFNS) from neutron-induced fission is a fundamental quantity for the behavior of nuclear reactors, and has been measured many times on a wide variety of nuclei and covering different ranges of incident and emitted neutron energies. However, results from past measurements are frequently called into question in modern nuclear data evaluations because of a lack of thorough experimental documentation and incomplete uncertainty analyses. The Chi-Nu experiment at Los Alamos National Laboratory was designed to produce high-precision measurements of the PFNS of major actinides over a wide range of incident and emitted neutron energies, and with the documentation and covariance analysis required to ensure that the results of this experiment maintain their impact long into the future, thereby avoiding this pitfall of past measurements. In this work we describe the Chi-Nu experiment along with summaries of the treatment of and methods developed to address two important components of the analysis of Chi-Nu data: random-coincidence backgrounds and MCNP simulations. Furthermore, we describe the first results for correlations not just between all data points collected on a single target nucleus, but also between all data points from separate Chi-Nu measurements on 235 U and 239 Pu. These correlations are important for accurately calculating ratios of the PFNS from one actinide to another, which are rare and can be informative for nuclear data evaluation efforts.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Resolved Resonance Region Analysis of 206 Pb, 207 Pb, and 208 Pb for Next Generation Lead-Cooled Fast Systems

Recently, great interest has been generated in using lead as a coolant for fast neutron systems and as a result it is important to investigate the ENDF/B-VIII.0 isotopic evaluations that comprise stable le ad. To this end, resonance parameters for 206 Pb, 207 Pb, and 208 Pb were re-evaluated because their resolved reso­nance regions extend beyond 0.5 MeV meaning resonance parameters used in reconstructing cross sections and elastic scattering angular distributions impact fast systems. The impact of resonance parameters is demon­strated by the differences between the evaluations in predicting experimental results from the fully modeled RPI Quasi-Differential Scattering Experiment via MCNPv6.2. In addition, MCNP KCODE calculations of lead-sensitive fast spectra critical benchmarks showed variations of keff on average of 400 pcm, caused solely from differences in elastic scattering angular distributions in 208 Pb. Re-evaluation entailed fitting data with the program SAMMY.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Kilopower Reactor Using Stirling TechnologY (KRUSTY) Component Critical Experiments

A series of critical experiments were conducted at the National Criticality Experiments Research Center (NCERC) in Nevada to evaluate the operational performance of a compact reactor that eventually will resemble the flight unit the National Aeronautics and Space Administration will use for deep space exploration. The results from the experiments are compared to preliminary results from computational models using MCNP and ENDF/B-7.1 neutron cross-section data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Critical Assemblies: Dragon Burst Assembly and Solution Assemblies

This work reviews the historical literature associated with the Dragon experiment and water boiler reactors operated at Los Alamos during the Manhattan Project. Frisch’s invited talk given at the American Nuclear Society’s Fast Burst Reactor Conference held at the University of New Mexico in Albuquerque, New Mexico, in 1969 is quoted. From the literature review, basic models for the Dragon experiment and for a water boiler–type assembly (aqueous homogeneous reactor) were created that can be used for conducting multiphysics simulations for criticality excursion studies. This methodology utilizes the coupled neutronic-hydrodynamic method to perform a time-dependent dynamic simulation of a criticality excursion. MCNP® was utilized to calculate important nuclear kinetic parameters that were incorporated into the models. Simulation results compare reasonably well with historic data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Designing Nuclear Fuels with a Multi-Principal Element Alloying Approach

Previous research has shown that multi-principal element alloys (MPEAs) using chromium, molybdenum, niobium, tantalum, titanium, vanadium, and zirconium can form stable body-centered-cubic (BCC) structures across a large temperature region (25°C to 1000°C). This is the same crystal structure as γ-uranium (U), which has shown desirable thermal and irradiation behavior in previous alloy fuel research. It is hypothesized then that the MPEA alloying approach can be used to produce a stable BCC uranium-bearing alloy and to retain its stability throughout anticipated operating regimes of power-producing reactors. Candidate elements were assessed using Monte Carlo N-Particle (MCNP) analysis to determine uranium densities necessary to make the alloy an economically viable fuel compared to conventional fuel forms. Following neutronic considerations, materials property databases and empirical predictors were used to determine the compositions with a high potential to form a BCC solid solution alloy. The final four alloys were MoNbTaU 2 , MoNbTiU 2 , NbTaTiU 2 , and NbTaVU 2 , which were cast using arc melting of raw elemental foils and chunks. Characterization of the fabricated alloys included scanning electron microscopy, X-ray diffraction, energy dispersive X-ray spectroscopy, and transmission electron microscopy. The results showed a two-phase system with a U-rich matrix phase surrounding the refractory precipitates. The U phase was found to contain varying concentrations of the alloying elements and was a BCC γ-U phase. These results warrant further research to identify ideal compositions for use as an advanced alloy fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Fuel Fabrication Specification Impact Analysis for NBSR LEU Conversion

As part of a national initiative to enhance nuclear security and reduce proliferation risks, significant efforts have been undertaken by the National Nuclear Security Administration Material Management and Minimization Office of Reactor Conversion Program to convert U.S. high performance research reactors (USHPRRs) from the use of highly enriched uranium (HEU) to low-enriched uranium (LEU), including the National Bureau of Standards Reactor (NBSR). The current plan is to procure LEU fuel assemblies from commercial fabricators according to fuel specifications tailored for each USHPRR. The analysis conducted at Brookhaven National Laboratory was part of an effort to identify the sources of uncertainty in the fuel specifications that may impact the performance of the NBSR core after its conversion and, in particular, to assess the range of acceptable tolerance limits from the perspective of core safety and reactor performance. Using the stochastic neutronics code MCNP 6.2, the variations in important NBSR neutronics characteristics were analyzed as a function of the specification parameters independently and in combination. The important NBSR specification parameters analyzed were the fuel isotopic composition, the amount of impurity content in cladding, the fuel plate thickness, and the fuel element 235U mass loading. The range of variation of each specification parameter was based on the technical specification limit or available as-fabricated assay data and uncertainties. The NBSR neutronics characteristics selected for analysis were the reactor reactivity characteristics at equilibrium and the equilibrium fuel cycle length. Results show that with variations in the fabrication parameters of the as-fabricated U-10Mo fuel within the specification limitations, the excess reactivity of the NBSR LEU core remains well below the 15% Δk/k technical specification limit, and the shutdown margin is always significantly greater than the required 0.68% Δk/k. This ensures that the NBSR can be operated safely and reliably shut down for all analyzed cases within the specified fabrication limits after the LEU conversion. In the prototypic case, the fuel cycle length was 1.5 days longer than the targeted 38.5 days. In a credible worst-case scenario, where all low-reactivity parameters were combined, the fuel cycle length was reduced to 35.5 days, which is still considered manageable for reactor operations. Variations in cycle length are primarily driven by changes in 235U loading, with other parameters having secondary effects.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

An Integral Experiment on Polyethylene Using Radiative Capture in Indium Foils in a High Flux D-D Neutron Generator

In this work, we report here the results of a measurement of the scattered versus unscattered neutron fluence on polyethylene determined via neutron activation of multiple natural indium foils from a deuterium-deuterium (D-D) neutron generator. The neutrons were produced by the High Flux Neutron Generator (HFNG) at the University of California, Berkeley, a specially designed source to maximize neutron flux on a sample while minimizing the total neutron yield. During the experiment, approximately 10 8 n/s were produced with the energies at the indium foils ranging from 2.2 to 2.8 MeV. Both the angle-integrated and the partial angle differential results are consistent with the predictions of the Monte Carlo N-Particle Transport (MCNP) code, using ENDF/B-VII.1. This supports shielding calculations in the fast energy region with high-density polyethylene.

42 ENGINEERING↗
DOE OSTI2021

Verification of Subcritical Limits in ANSI/ANS-8.1-2014

The first edition of the current American National Standards Institute (ANSI)/American Nuclear Society (ANS) standard ANSI/ANS-8.1-2014, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors,” was published in 1964 as ASA N6.1-1964, “Safety Standard for Operations with Fissionable Materials Outside Reactors.” In 1969, that standard was revised as ANSI N16.1-1969, “Nuclear Criticality Safety in Operations with Fissionable Materials Outside Reactors.” ANSI N16.1-1969 includes a variety of subcritical limits (SCLs) for uniform aqueous solutions and metals containing fissile nuclides of 233 U, 235 U, and 239 Pu. Furthermore, SCLs are also included for uranium-water lattices. In the 1983 version of ANSI/ANS-8.1 (a revision of ANSI N16.1-1975), the suite of SCLs in the standard was expanded to include 235 U enrichment limits for homogeneous uranium-water mixtures and dry/damp oxides, uniform aqueous solutions of low-enriched uranium, and uniform aqueous mixtures of Pu(NO 3 ) 4 containing 240Pu in addition to the SCLs included in ANSI N16.1-1969. The SCLs have changed little in subsequent revisions (ANSI/ANS-8.1-1998 and ANSI/ANS-8.1-2014). The ANSI/ANS-8.1-2014 standard is currently being revised to include new SCLs (uranium metal and compounds with enrichments up to 20 wt% 235 U) and possible updates to the current SCLs already in the standard, although these new/updated SCLs will not be available to the nuclear criticality safety (NCS) community for a number of years. The original bases for these SCLs were documented in papers in journals such as Nuclear Science and Engineering and in ANS meeting transactions; however, these bases are ambiguous enough that sites and regulators in the United States have not widely endorsed them for safety purposes. Overall, the purpose of this paper is to present the results of a comparison study for the SCLs in the ANSI/ANS-8.1-2014 standard using modern codes (SCALE and MCNP) and cross sections (ENDF/B-VIII.0) to provide some assurance about their quality (bias and bias uncertainty) for use in NCS applications and for consideration by the ANS-8.1 Working Group as a reference for future revisions.

42 ENGINEERING↗