Neutron-state entanglement with overlapping paths
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Results for “nuclear instruments”
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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Chemical and elemental mapping of spent nuclear fuel sections by soft X-ray spectromicroscopy
Soft X-ray spectromicroscopy at the O K -edge, U N 4,5 -edges and Ce M 4,5 -edges has been performed on focused ion beam sections of spent nuclear fuel for the first time, yielding chemical information on the sub-micrometer scale. To analyze these data, a modification to non-negative matrix factorization (NMF) was developed, in which the data are no longer required to be non-negative, but the non-negativity of the spectral components and fit coefficients is largely preserved. The modified NMF method was utilized at the O K -edge to distinguish between two components, one present in the bulk of the sample similar to UO 2 and one present at the interface of the sample which is a hyperstoichiometric UO 2+ x species. The species maps are consistent with a model of a thin layer of UO 2+ x over the entire sample, which is likely explained by oxidation after focused ion beam (FIB) sectioning. In addition to the uranium oxide bulk of the sample, Ce measurements were also performed to investigate the oxidation state of that fission product, which is the subject of considerable interest. Analysis of the Ce spectra shows that Ce is in a predominantly trivalent state, with a possible contribution from tetravalent Ce. Atom probe analysis was performed to provide confirmation of the presence and localization of Ce in the spent fuel.
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Sandia 7uPCX Critical Experiments Exploring the Effects of Fuel-to-Water Ratio Variations
The Sandia Critical Experiments (SCX) Program provides a specialized facility for performing water moderated and reflected critical experiments with UO 2 fuel rod arrays. A history of safe reactor operations and flexibility in reactor core configuration has resulted in the completion of several benchmark critical experiment evaluations that are published in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. The LEUCOMP-THERM-078 and LEU-COMP-THERM-080 evaluations from the handbook provide similar cases for reference. The set of experiments described here were performed using the Seven Percent Critical Experiment (7uPCX) fuel to measure the effects of decreasing the fuel-to-water volume ratio on the critical array size. This was accomplished by using fuel loading patterns to effectively increase the pitch of the fuel arrays in the assembly. The fuel rod pitch variations provided assembly configurations that ranged from strongly undermoderated to slightly overmoderated.
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System-Independent X-Ray Characterization
We have developed a method for image reconstruction and analysis of dual-energy X-ray scans that has been demonstrated to yield a physically intuitive characterization of materials (by electron density and effective atomic number) in a system-independent way. Our methodology stands in contrast to common practice of reconstructing images to yield linear attenuation coefficients, which depend strongly on system spectral response. At the core of our method, a technique called dual-energy decomposition is used to generate spectral energy-independent characterizations of the attenuation properties of a material. Inputs to this decomposition include system spectral response estimates and dual-energy X-ray projection data. X-ray cross-section tables are used to convert decomposition results to physical feature estimates (electron density, effective atomic number). Our method currently works best for materials with atomic numbers below 22 (titanium). In prior work for DHS, we demonstrated in a controlled, no-clutter environment that the decomposition technique could be applied to scans of a well-characterized materials on multiple systems with different X-ray spectral responses. In cases where LAC values showed a standard deviation of as much as 20% of mean value, application of our method yielded results with standard deviations as low as 1–3% of mean value.
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(U) Endpoint Energy and Scintillator Geometry Effects on the Swank Factor, Quantum Efficiency, and DQE(0) for High-Energy Radiography
High-energy radiography is used to image imploding materials to better understand weapon design and performance. The Enhanced Capabilities for Subcritical Experiments (ECSE) is a new facility proposal that aims to collect radiographic images of nuclear material. As part of the facility design, we aim to optimize data quality and radiographic performance by investigating different detector designs. We addressed design decisions by simulating different scintillator designs using M CNP6 and calculated the quantum efficiency, Swank factor, and detective quantum efficiency at zero-frequency (DQE(O)). First, we investigated the effects of the x-ray endpoint energy on the detector response. We generated three x-ray spectra with endpoint energies of 20 MeV, 22 MeV, and 24 MeV. The three raw spectra were filtered using the Dual-Axis Radiographic Hydrodynamic Test facility (DARHT) line-of-sight materials along with 1) flat field filtration using 7 cm tungsten alloy in the bullnose and 2) an FTO with 4 plates. The scatter-to-direct radiation ratio was much larger for FTO with 4 plates (50-60%) compared to the flat field with 7 cm of bullnose material (8%). We used these filtered spectra to calculate the detector response for the original DARHT segmented scintillator design. We found that the endpoint energy and filtration had a negligible effect on the detector response. The quantum efficiency was 0.63, the Swank factor was 0.84, and the DQE(O) was 0.53 for all six spectra.
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NDSE Static Series V & VI Test Results
The purpose for the Neutron-Diagnosed Subcritical Experiment (NDSE) Static Test Series at the Nevada National Security Site (NNSS) Area 11 is to develop and validate the capability to make precise and accurate measurements of k eff for Special Nuclear Material (SNM) targets with k eff ~ 0.95.1 The series has now completed six sets of measurements using the deuterium-tritium (DT) Dense Plasma Focus (DPF) neutron source, with a Rocky-Flats-Shells Highly Enriched Uranium (RF HEU) target, to measure the gamma die-away to this purpose. The last of these series, Series V and VI, are the subject of this report. Both series were performed on a target comprised of the Object IV HEU + CH 2 configuration, which is identical to Object II but with smaller DPF collimators that limit the incident neutron flux to the HEU radius.
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Tomographic Image Reconstruction for the Parallel-Slit Ring Collimator Fast Neutron Emission Tomography System
For the past three years, Oak Ridge National Laboratory (ORNL) has been developing a passive fast-neutron emission tomography capability. The goal of this development is the ability to quantify the neutron source strength of individual fuel pins (rods) in spent nuclear fuel assemblies. Such a system could be used to measure the burnup of each fuel pin in a spent fuel assembly in order to take burnup credit when loading dry storage casks or to count individual fuel pins in spent fuel assemblies for safeguards purposes. At present, a laboratory prototype imager is under construction. The purpose of this prototype is to demonstrate imaging capability sufficient to resolve individual fuel pins in spent fuel assemblies. This report documents the development of the iterative reconstruction code used to perform tomographic image reconstruction, the imager response calculation used by the reconstruction code, and the results of reconstructions of simulated tomographic imaging measurements for the prototype imager design.
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Monolithic Spatial Heterodyne Raman Spectrometers (mSHRS): Toward highly miniature chemical sensors for deployment on moving platforms (Final Report, Phase I)
This report covers phase 1 (year 1) of the mSHRS project. Our ultimate long-term goal is to develop highly miniature Raman analyzers, known as spatial heterodyne Raman spectrometers (SHRS), that are deployable as in-situ chemical sensors on moving platforms. During Phase I (FY19/20) funding, our team focused our efforts solely on the key innovation needed toward realizing that goal: the development of the monolithic interferometer comprising the beam splitter and gratings with the appropriate design parameters and tolerances to enable Raman interferometry. We successfully developed and characterized the performance of the first monolithic SHRS devices including a monolithic version that incorporates field-widened optics. Our solid-state design approach integrated all optical components as a single monolithic unit for maintaining precise optical alignment during sensor operation, a requirement for operating in a vibration prone environment. This project has two principle activities: (1) Design and fabricate monolithic SHRS devices (i.e. monolithic device task) and (2) characterize the mSHRS devices in terms of performance (i.e. the Monolithic SHRS characterization task).
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An Evaluation of a Neutron Time Correlated Interrogation Method for Measurement of Fissile Content in Research Reactor Spent Fuel Assemblies
There are many research reactors worldwide that have been, or are being converted from the use of high enrichment uranium (HEU) to low enrichment uranium (LEU, < 20% enriched). The verification of fissile content and initial enrichment of the spent fuel is needed for the effective safeguards of the fuel. The advanced experimental fuel counter (AEFC) was developed for the measurement of spent fuel rods and assemblies from research reactors for safeguards verification. This measurement system contains components for active neutron interrogation, passive neutron totals counting, neutron coincidence counting, and gross gamma-ray counting. This report presents the first application of the time correlated interrogation technique for the measurement of the 235 U content in research reactor spent fuel assemblies. The technique, called time correlated induced fission (TCIF), uses a 252 Cf neutron source to irradiate the fuel assembly, and the subsequent induced fission events in the fissile material are measured by coincidence counting. The doubles rates are enhanced by having the neutron trigger events from both the 252 Cf source and the induced fission neutrons in the same time gate in the coincidence analysis. The average neutrons per fission of the 252 Cf source is 3.76 and the induced fission neutrons for 235 U is 2.44, so the number of neutrons that are produced is higher than for random neutron interrogation. This high effective neutron number increases the multiplicity counting rates and reduces the statistical error. The background coincidence counts from the 252 Cf are reduced by the water in the sample cavity and the polyethylene surrounding the 3 He detector tubes. This method of active neutron interrogation has been applied to the measurement of spent research reactor (IRT) fuel assemblies. The advanced experimental fuel counter (AEFC) was used to compare the TCIF method with the typically used AmLi neutron interrogation source that emits neutrons that are random in time. The statistical uncertainty for the use of the random neutron source (AmLi) and the time correlated source ( 252 Cf) for spent fuel interrogations was evaluated.
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Total Measurement Uncertainty in Neutron Coincidence Multiplicity Analysis
Neutron multiplicity counting is the most commonly used nondestructive assay technique for determining the plutonium mass within containers of scrap PuO 2 or mixed oxide (MOX). In multiplicity analysis, the 240 Pu eff mass, leakage multiplication, and alpha ratio (the ratio of [α, n]-to-spontaneous fission neutron production) are the three primary unknown sample properties. They must be determined simultaneously. To solve for these three unknowns in a multiplicity assay, three measured values are needed: the singles, doubles, and triples neutron count rates. While the analysis is limited to solving for three unknowns, there are many additional factors that impact the observed count rates and contribute to the measurement uncertainty. In this study we investigate the various uncertainty contributors for the multiplicity analysis through a combination of traditional uncertainty propagation techniques supplemented by Monte Carlo simulations to address the dependences not explicitly expressed by the point source model. Uncertainties arising from counting statistics, calibration parameters, calibration method, nuclear data, and various material characteristics (isotopic abundances, chemical form, density, and impurities) are considered. A Total Measurement Uncertainty (TMU) estimate is then developed from these uncertainty contributors. This study is confined to multiplicity analysis of items commonly encountered in international safeguards applications. That is, the study focused on Pu oxides and MOX materials for the masses ranging up to 4000 grams total Pu. Multiplicity measurements were simulated using MCNP V6 based on the Plutonium Scrap Multiplicity Counter (PSMC), Epithermal Multiplicity Counter (ENMC), Pyrochemical Multiplicity Counter, and Large Epithermal Multiplicity Counter (LEMC) for this study; however, this report focuses on the parameterization of the uncertainties for the PSMC. The performance differences between the PSMC and the other multiplicity counting systems are relatively small, primarily manifesting in the impact on measurement precision so that the evaluation developed for the PSMC can be applied to the other multiplicity counting systems. Finally an analysis tool, the Multiplicity TMU Estimator, was developed from this study to serve as an aid for evaluation of the total measurement uncertainty of multiplicity assay results obtained from the commonly used INCC acquisition and analysis software.
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Effects of 18 O Exchange on Neutron Emission Rates of Aging 238 PuO 2 LWRHUs
The Lightweight Radioisotope Heater Unit (LWRHU) provides about 1 Watt of thermal power to critical electrical and mechanical systems in the extreme temperatures of space, maintaining the operating temperature of the spacecraft’s scientific and general operating equipment. Los Alamos National Laboratory (LANL) designed and tested the LWRHU in the 1970s through early 1980s. The first production campaign at LANL was in the early 1980s in support of the Galileo mission in which over 100 LWRHUs were manufactured. In the mid-1990s, almost 200 LWRHUs were manufactured in support of the Cassini mission. After the primary Cassini manufacturing campaign, the Department of Energy (DOE) approved the fabrication of more than 10 additional LWRHUs as spares in 1998. The LWRHU assembly contains the following components: a fuel pellet, a vented capsule, a pyrolytic graphite insulator, and a fine-weave pierced fabric (FWPF) graphite aeroshell. The fuel pellet is a hot-pressed 238 PuO 2 cylindrical pellet that is sintered to create a ceramic pellet. The pellet is encapsulated in a platinum-rhodium encapsulation. The welded fueled clad is placed inside a pyrolytic graphite thermal insulation to protect the fuel from the heat produced in reentry events. The exterior layer of the LWRHU assembly is the FWPF aeroshell. Before sintering, PuO 2 granules are heated in a furnace at temperatures greater than or equal to 700°C under an enriched 16 O atmosphere in order to reduce trace quantities of 18 O. Because 18 O undergoes an (α,n) reaction, 18 O needs to be removed as much as possible to meet radiation emission requirements set by the Jet Propulsion Laboratory (JPL) and the National Aeronautics and Space Administration (NASA), as well as other agencies. The relative abundances of the oxygen isotopes are indicated by neutron emission rate (NER) measurements. LWRHUs are vented with a sintered platinum frit vent (George 1986). During extraterrestrial operation, the vent in the cladding allows helium from the decay of plutonium to escape and relieves pressure from the system while preventing the solids from escaping (Rinehart 1996, Tate 1982 & 1985). This vent is protected during manufacture by a platinum-30 rhodium cover. The capsule vent is activated by milling a 0.025-inch diameter hole through the protective cover to a nominal depth of 0.015 inches, generally immediately before the LWRHU is loaded into the aeroshell (George 1986). Prior to long-term storage, LWRHUs are vented and thus potentially allow infiltration of 18 O from the ambient atmosphere. Although NER decreases with age due to the reduction in emitted alpha particles, it might be expected that NER reduction is mitigated by the exchange of 18 O. However, this is not necessarily the case and no trend was observed relating NER to age other than the natural decay of 238 Pu (Mulford 2021). The most probable and dominant factor for determining specific NER was manufacturing discrepancies between individual units. It is important to understand the lack of observed 16 O- 18 O exchange and to compare LWRHU pellets as much as possible.
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Independent Gamma Spectroscopy Analysis of a Plutonium Oxide Sample
Independent gamma spectroscopy data analysis of a plutonium oxide sample was requested on July 23, 2021. The primary request was to assess the Pu-239 activity/mass of the sample using previously collected gamma spectral data. Using the provided gamma spectral analysis report and spectral files, an independent evaluation of the data was conducted without any prior knowledge of the isotopic activity/mass of the sample.
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Fixed-Energy Response-Function Analysis with Multi-Efficiency (FRAM) Software, Version 6
Researchers at Los Alamos National Laboratory have developed FixedEnergy Response- Function Analysis with Multiple Efficiency (FRAM), a powerful software tool for gamma-ray spectrometry isotopic analysis. This tool leverages Los Alamos expertise in nuclear safeguards, radiation detection systems, and related analysis techniques. A household name in the non-destructive assay community for over two decades, FRAM determines a special nuclear material source's isotopic composition by analyzing gamma-ray spectra. Los Alamos researchers have released the latest version of FRAM that incorporates several updates to the 2013 major release of FRAM, Version 5.
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233 U(n,$\gamma$) measurements at LANSCE [Slides]
A new measurement at LANSCE combined DANCE and NEUANCE at the end of 2020 and 2021. The 233 U material was provided by Oak Ridge National Laboratory (December 2020).Two samples of 10 mg and 20 mg of 233 U have been prepared at LANL by Evelyn M. Bond (December 2020). The experiment was performed in December 2020 and June/July 2021. Data analysis is ongoing to extract capture data knowing fission. LANL is looking forward to delivering the capture/fission ratio from (1-300) keV by the end of the FY2022.
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Canister Helium Leak Detection Based on Temperature Information Using RAMM (Final CRADA Report)
The performance of RAMM-TM in detecting gas leakage from a canister with a small, simulated chloride-induced stress corrosion crack has been demonstrated in a series of experiments conducted at CRIEPI by using a 1/4.5- scale model cask. Leakage of both helium and air from a canister was detected within hours after the start of the leakage. The change in surface temperatures at the top and bottom of the canister (ΔTBT) during gas leakage (depressurization) triggered automatic alarms, providing a sound basis for early detection of gas leakage from the canister. This methodology would enable consequence management through the implementation of mitigatory actions to continue effective aging management, to reduce risks to public safety and health, and to protect the environment.
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High-Energy X-ray Diffraction Microscopy for Nuclear Forensics FY2022 Project Report
Morphological information on nuclear material has been identified using visible light and scanning electron microscopy. These identify qualitative differences in particle morphology. Three-dimensional imaging of materials through alternating scanning electron microscopy imaging and focused ion beam milling has also been used. Unfortunately, these techniques are time- and labor-intensive, with significant sample preparation required and lengthy analysis times. Further, the resulting 3D images are qualitative, require manual identification, and do not capture statistically-representative populations. High energy X-ray 3D imaging using a direct-beam or diffracted-beam (High-Energy Diffraction Microscopy) have been developed at the Advanced Photon Source and can produce quantitative information on grains (phase, location, etc.) and pores (size distribution, sphericity) in a material. These techniques require only minutes to characterize a sample volume and are non-destructive, thus suitable for a wide range of existing samples and for confirmatory analyses to be carried out using conventional microscopy techniques. In this first year of the project, all uranium oxide samples were synthesized and characterized using conventional analyses by the analytical chemistry laboratory. Conventional analysis methods included powder x-ray diffraction, scanning electron microscopy, impurity analysis via inductively coupled plasma mass spectrometry, and infrared spectroscopy. Impurity analysis shows a drop in boron content from UO 3 to the lowest U 3 O 8 calcination temperature, but otherwise no appreciable difference in any sample. Analysis of diffraction data shows a flip of peaks from UO 3 dominated for the 600 °C calcined sample to U 3 O 8 dominated at 700 °C and 800 °C. Analysis of scanning electron microscopy images shows that with increased calcination temperature the size distribution of particles seems to increase and broaden. Both of these last findings are in line with previously published data, though this work used significantly fewer particles to simply show similar trends instead of getting truly quantitative particle analysis. Infrared analysis similarly shows ingrowth of U 3 O 8 as calcination temperature is increased, along with depression of peaks associated with UO 3 and water. Samples were prepared for analysis at the Advanced Photon Source at beamline 1-ID. It is anticipated that analysis will occur in November of 2022. AI/ML techniques to de-noise data coming out of 1-ID during the analyses was also developed during this time using previously gathered data. Preliminary results using a self-supervision technique called Noise2Selfshow good de-noising of data. Once the uranium oxide samples are analyzed, real data will be used to test the de-noising and other AI/ML techniques that may be developed in the second year of the project.
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Density Measurement of NaCl-MgCl 2 -PuCl 3 and NaCl-UCl 3 -PuCl 3 Molten Salt Systems by Neutron Radiographic Dilatometry
A novel neutron radiographic technique was developed and implemented to measure the change in volume as a function of temperature of select plutonium-containing liquid chloride mixtures from melt to 1250 K. These measurements were performed at Los Alamos National Laboratory by members of the molten salt research group under support from Gateway for Accelerated Innovation in Nuclear (GAIN) voucher NE-21-25117: “Density Measurements of Plutonium Bearing Salts via Neutron Beam Dilatometry”. This technical report contains a description of the experimental method and a summary of the results. Initial characterization of the {NaCl, MgCl 2 , UCl 3 , PuCl 3 } salts used in this experiment includes a combination of x-ray diffraction, differential scanning calorimetry, gamma ray spectroscopy, and inductively coupled plasma mass spectrometry. Four samples of NaCl + MgCl 2 + x mol% PuCl 3 and four samples of NaCl + UCl 3 + x mol% PuCl 3 were prepared at the Plutonium Facility, shipped to the Los Alamos Neutron Science Center (LANSCE), and radiographed with the density via neutron radiography (DvNR) apparatus. Density data in the form of meniscus heights were recorded as a function of temperature. The results show a linear dependence of density with temperature and a third order polynomial dependence of density with composition as a function of PuCl 3 addition in the corresponding halide salt mixture. Additions of PuCl 3 result in a significant increase in density of the salt mixture and a slight reduction in the coefficient of volumetric expansion. These results represent a first measurement of the density of plutonium chloride salt mixtures and, in combination with additional measurements, may lead to a generalized description of the density of such salt systems for arbitrary compositions and temperature.
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Photogrammetry: Develop methods for detecting deviations from expected geometry of components in a glovebox environment (Lawrence Livermore National Laboratory Aging and Lifetimes Program FY23, Milestone 8651, Grading Criterion #4)
Measurements in glovebox environments are challenging to conduct. Currently deployed techniques are time consuming and provide limited information. These limitations make current disposition and future assessments of products challenging. In FY23 we developed methods for detecting deviations of components in a glovebox environment that address these shortcomings by rapidly collecting information-dense measurements. Conducted tests demonstrate that camera pose (position and orientation) can be derived for images utilized in the photogrammetry process. This information, in conjunction with common image processing techniques, enables automated detection and size estimation of surface features. First, image processing is used to identify marks, or localized regions, that standout from the surrounding area. Then, the physical size of the region can be estimated because scale can be determined for photogrammetry images. Development of this methods shows that photogrammetry can identify measurable deviations from expected geometries. Although measurement error when using this technique through a glovebox window still needs to be reduced, this capability shows promise for eliminating or reducing the use of tedious manual measurements.
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