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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Using FENOSOL{sup TM} Phenolic Foam as a Shock Absorber for Type B Packaging - 17256

Type B and fissile material transportation packages are designed to withstand deformation from dynamic forces experienced during impacts in hypothetical accident conditions. A critical aspect is to ensure the containment system is not breached. The package design normally includes impact limiters made up of a shell filled with energy-absorbing material. Various types of materials are used as energy absorbers, including wood, foam, and honeycomb structures. Synthetic foam is an ideal material for impact limiters; it has isotropic material properties, it acts as a heat insulator during fire conditions, and it maintains structural integrity at high temperatures. In addition, the density of the foam may be controlled during the manufacturing process, allowing the cask designer flexibility to select the right density for each application. The foam properties can be used as inputs for numerical simulations of dynamic impacts. These simulations are a key component of the safety analysis report for Type B transportation casks. Robatel has recently invested in equipment to produce FENOSOL{sup TM} phenolic foam in our factory for use in our cask designs. This acquisition was made jointly between Robatel and the CEA, with Robatel ensuring the industrialization of the product. In the past decade, this foam was successfully integrated into more than 10 European cask designs, and there has been extensive in-laboratory characterization of its properties. FENOSOL{sup TM} has already been accepted by the French Nuclear Safety Authority (ASN) as a shock absorbing material for use in cask designs. Since the acquisition of the phenolic foam production process by Robatel, a new research and development program has been launched to provide more accurate information to our clients, extend the range of application of this foam, and to improve its properties. Robatel has built an extensive database of material characteristics through the use of a universal mechanical tester, a scanning electron microscope, and a nuclear magnetic resonance spectrometer. In addition, Robatel has successfully implemented FENOSOL{sup TM} in two recent cask designs. This included scale model drop testing of a 9 meter drop in three orientations, and a pin puncture drop test. Robatel is in the process of producing FENOSOL{sup TM} samples for testing according to ASTM standards. Various densities will be included in the study. The testing will include duro-meter hardness, specific heat, thermal conductivity, compressive strength and modulus, and tensile strength. Once the testing is complete, Robatel will compile the data in the form of product specification sheets for each density. These information sheets and test reports will be used to validate FENOSOL{sup TM} for use in future impact limiters for the North American Type B cask market. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Full Length Assembly Testing in PELICAN (Final Report)

In support of the development of the U.S. Department of Energy (DOE) Versatile Test Reactor (VTR), a thermal hydraulics test facility was constructed to generate experimental measurement of the pressure drop across a single full-scale assembly containing prototypic axial reflectors, fuel, and plena components. Constructed and operated at Argonne National Laboratory, the Pressure drop Experimental Loop for Investigations of Core Assemblies in Nuclear reactors (PELICAN) facility was designed to achieve hydraulic conditions identical to those anticipated for a full-scale fuel assembly located in the VTR core in the region with the highest flow rate. Using water as surrogate for liquid sodium, the flow loop was operated at elevated temperatures and pressures to match the thermophysical properties of liquid sodium and ensure matching Reynolds and Euler numbers. The measurement objectives for data generated from this test facility was driven primarily by the validation needs for code calculations and simulations of the reference VTR core. These objectives focused on the need to validate pressure drop results across the various segments of the fuel assembly as they relate directly to the pumping power and safety of the reactor. Presented in this report are experimental results and analytical comparisons based on testing of a full-length assembly in PELICAN. Housed within a hexagonal test section extending 3.4 m in length, the tested assembly features a prototypic lower reflector, grid plates, wire-wrapped rod bundle, upper reflector, and exit region. The rod bundle extends over 1.5 m in length and contains 217 individual wire-wrapped rods with dimensions that best reflect the reference VTR design. The as-tested bundle assembly was fabricated using 316 stainless steel 0.25-inch (6.35-mm) diameter rods wrapped with 0.04-inch (1.016-mm) diameter wire at a helical pitch of 10.51 inch (26.6 cm). Details of the method for in-house wire-wrapping, assembly, and installation are provided later in this report. Experimental measurements of pressure drop at 19 positions along the test assembly were recorded for a range of flow conditions, with special attention paid to key locations within the assembly, including component inlet and outlet, transition, and wire-wrapped rod bundle regions. Testing conditions were based on 110°C water with flow rates ranging from 50 to 450 GPM (3 to 27 kg/s) at the inlet of the test assembly generating Reynolds numbers and velocities up to ~8.0×10 4 and ~7.8 m/s, respectively, within the rod bundle region. Non-dimensional values for the friction factor were then calculated based on these experimental measurements and compared against those predicted by various analytical correlations available from open literature. Predictions by the upgraded Cheng and Todreas, Rehme, and Novendstern correlations fell within 4% to those values measured experimentally.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Advanced Instrumentation for Transient Reactor Testing

Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and design increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Thermal-Hydraulic Results for the Aboveground Configuration of a Dry Cask Simulator

The thermal performance of a commercial nuclear spent fuel dry storage cask is evaluated through detailed analytical modeling. These modeling efforts are performed by the vendor to demonstrate the performance as well as regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). The majority of commercial dry storage casks in use today are aboveground. Cooling of the assemblies located inside the sealed canister is achieved by naturally-convective air flowing in the bottom of the cask and exiting out the top. Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating design and performance models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation efforts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. Thus the enhanced performance of modern dry storage casks cannot be fully validated using previous studies.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

Drying Spent Magnox Fuel for Interim Dry Storage

Spent Magnox fuel is usually stored under water for shielding and cooling following use in-reactor, but the storage water can react with Magnox clad fuel and cause corrosion and forms a sludge which generates a secondary waste product. One option to prevent unwanted corrosion and avoid expensive waste management is to dry the wet fuel. This project aims to test a drying process which will convert wet and corroded Magnox fuel to a stable form to demonstrate viability for conversion from wet to dry storage. Samples of corroded Magnox supplied by National Nuclear Laboratory. Corrosion product retrieved by various methods including dry scraping, ultrasonic treatment to form sludge and sludge drying to isolate suspension. TGA undertaken on Sample 1 and Sample 2, with brucite and hydromagnesite for comparison. Both samples showed ∼10% mass loss up to 250 deg. C attributable to water held in sample. Mass loss from dehydroxylation which liberates further water in both samples. TGA data plotted alongside chemically pure brucite and hydromagnesite shows expected similarities/differences between Sample 1(mostly brucite) and Sample 2 (brucite/hydromagnesite mixture). Following treatment, dried samples observed to slowly gain mass in ambient conditions. Presumed to be hygroscopic absorption of air water vapour as effect is reversible, with similar effect observed for hydromagnesite and brucite. Etched Magnox vacuum dried for 3.5 h at 90 deg. C to test for reaction with residual water - some discoloration but no noticeable corrosion observed. Drying etched Magnox metal to observe surface effects/corrosion during drying process.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Evaluating Age Effects on HEPA Filter Performance

Nuclear grade high efficiency particulate air (HEPA) filters are routinely employed as the final control element in confinement ventilation systems. The traditional HEPA filter utilizes a pleated filter medium of fibrous glass that has been produced in a manner analogous to manufacturing paper. These filters are the physically weakest element in a containment ventilation system. Containment ventilation systems are designed to condition the air-stream (temperature, relative humidity, moisture, particle size distribution of aerosols, and aerosol concentration) to levels/conditions that fibrous glass HEPA filters can handle. Approximately 6000 nuclear grade HEPA filters are purchased each year for use in the DOE/NNSA complex. Additional filters are purchased for use in commercial nuclear power generation and ventilation systems of other nuclear applications. These filters are designed and tested in accordance with the American Society of Mechanical Engineers (ASME) AG-1 Code Nuclear Air and Gas Treatment. This code establishes design requirements and qualification procedures. Qualification testing of axial flow filters is described in paragraphs 5000 and following of AG-1. The series of qualification tests are intended to describe the operating envelope of HEPA filters and verify the reliability of a filter design. AG-1 qualification also confirms the pedigree of components used in manufacture for that filter design. The collective set of qualification tests establishes the assumed operating envelope for AG-1 filters. The effects of age on HEPA filter performance have been a concern for at least two decades. A combination of physical evaluation of aged filter media and a review of filter failure records resulted in Appendix C of the Nuclear Air Cleaning Handbook (NACH), Determination of HEPA Filter Life. This states that HEPA filters are not to be used beyond 10 years from the date of manufacture. Additionally, filters put into service in locations where they may become wet are to be taken out of service five years from date of manufacture. Filters that have become wet are to be taken out of service immediately. Finally, it should also be pointed out that filters held in Level B storage are subject to the same 10-year service life. Very limited bench-scale testing in the 1990's raised concern that aging HEPA filters do not have the strength to withstand an accident scenario. In May of 1999, the DNFSB released a Technical Report 23 entitled HEPA Filters Used in the Department of Energy's Hazardous Facilities. This report expressed concerns for the potential vulnerability of HEPA filters in vital safety systems. Concerns and uncertainty associated with degradation of HEPA filter performance over time led the DOE sites to limit HEPA filter service life to 10 years from the date of manufacture or five years in cases where filters may become wet. Establishment of a conservative service life needs to be based on data from a structured series of tests comparing the performance envelope of new and aged full-scale filters. A service life that is insufficiently conservative endangers workers and the public. One that is excessively conservative costs can cause hundreds of otherwise unnecessary filter changes. This increases exposure of employees, disrupts facility operations, and increase disposal costs by millions of dollars annually. Conclusive data are needed to resolve uncertainty associated with the damaging effects of aging on durability of HEPA filters. Therefore, this study needs to provide a sufficient body of evidence to allow DOE and site professionals to make prudent decisions. An updated version of ASME AG-1 Section FC included a non-mandatory appendix that would include the DOE guidance for service life of HEPA filters. Balloting of this revision generated significant resistance to inclusion of any kind of service life language citing the lack of supporting experimental data. Efforts to resolve negative votes failed and resistance to acceptance of the language increased. The result has been to not include service language in Section FC and requests to remove it from sections that already include such language. The recent activity of the ASME AG-1 Committee on Nuclear Air and Gas Treatment effectively de-emphasizes following DoE's service life guidelines. The DOE Office of Nuclear Safety under its NSR and D program provided the Mississippi State University Institute for Clean Energy Technology (ICET) finding to initiate a study to determine the effects of aging on AG-1 axial flow filters just prior to actions taken within the AG-1 committee regarding service life. The NSR and D funding facilitated timely development of a national technical working group comprised of members from the regulatory community, DOE complex, and commercial power to resolve uncertainty associated with effects of aging. The Department of Energy Office of Environmental Management (DOE-EM) has been funding the ICET to evaluate the performance envelope for AG-1 HEPA filters. ICET is one of DOE-EM's qualified suppliers. Synergy has been achieved by associating the NSR and D aged filter activities to testing activities evaluating the performance envelopes of separator and separator-less AG-1 HEPA filters funded by DOE-EM. The DOE-EM funded activities establish baseline data for the aged filter study and future EM funding will provide long term support for testing additional aged filters. EM funding of new filters includes analytical evaluation of media tensile strength by thermal analysis to establish baseline values and determine mass loss of moisture resistance and acrylic binders due to use. All of these values can be used for comparison of test results for aged filters. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Toward shell model interactions with credible uncertainties

Background: The nuclear shell model is a powerful framework for predicting nuclear structure observables, but relies on interaction matrix elements fit to experimental data as its inputs. Extending the shell model's applicability, particularly toward dripline nuclei, requires efficient fitting methods and credible uncertainty quantification. Traditional approaches face computational challenges and may underestimate uncertainties. Purpose: We develop and test a framework combining eigenvector continuation and Markov chain Monte Carlo to efficiently fit shell model interaction matrix elements and quantify their uncertainties. Methods: Eigenvector continuation is used to emulate shell model calculations, reducing computational costs. The emulator enables Markov chain Monte Carlo sampling to optimize interaction matrix elements and rigorously assess parametric uncertainties. Here, the framework is benchmarked using the USDB interaction in the 𝑠⁢𝑑 shell. Results: The emulator reproduces the USDB interaction with negligible error, validating its use in shell model fitting applications. However, we find that to obtain credible predictive intervals, the model defect of the shell model itself, rather than experimental or emulator error, must be taken into account in order to obtain credible uncertainties. Conclusions: The proposed framework provides an efficient and rigorous approach for fitting shell model interactions and quantifying uncertainties. Further, the normality assumption used in the past appears sufficient to describe the distribution of interaction matrix elements. However, it is crucial to account for model correlations to avoid underestimating uncertainties.

Nuclear forces↗
DOE OSTI2024

MSR Transient Simulation and MSRE Transient Benchmark with SAM and SPECTRA

In recent years, there has been renewed interest in molten salt reactors (MSRs) for their potential advantages compared to reactors that rely on solid fuel. In response to such interest, many methods and codes have been developed to capture the unique features of MSRs. Among them, SPECTRA and SAM are two system analysis codes that have been enhanced to include MSR-specific modeling capabilities, including delayed neutron precursor drift and modified point kinetics equations.This paper discusses the efforts taken to verify and validate these features. A standard MSR system test problem was developed to verify and demonstrate the capability of SPECTRA and SAM on the MSR transient simulation. Sixteen transients were simulated. Further, the results obtained from SPECTRA and SAM show good agreement. The Molten Salt Reactor Experiment transient experiments were reviewed and selected to validate the SPECTRA and SAM codes. The experiments included pump startup and coastdown tests at zero power, reactivity insertion tests at different power levels, frequency tests, and a natural convection test. The simulation results from SPECTRA and SAM show good agreement with the experimental data.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

MCNP Simulation of Discrete Gamma-ray Spectra for PGNAA Applications

Prompt Gamma Neutron Activation (PGNAA) systems use the neutron-induced gamma-ray spectra from test objects in order to identify their compositions. The U.S. Army in particular uses these systems to identify the contents of field-recovered munitions in order to determine the presence of chemical warfare materiel (CWM) or explosives. These field-recovered munitions can have a variety of different fills, including experimental fills, have a large range of sizes, and be assessed in different field conditions. All these variations can have a significant effect on the neutron-induced gamma-ray spectrum and therefore a large impact on automatic fill identification algorithms. In order to develop robust identification algorithms, parametric studies of the effects of munition size, fill composition, and assessment conditions are very useful but time-consuming and expensive in the laboratory. Monte Carlo modeling of the system response to these variations is a very attractive alternative, provided the modeling accurately reflects real gamma spectra and can be performed quickly and inexpensively when compared to experiment. At Idaho National Laboratory we have used the Monte Carlo N-Particle (MCNP6) code [2] to calculate the response of the Portable Isotopic Neutron Spectroscopy (PINS) system to a number of simulated chemical warfare munitions as well as performed laboratory measurements with the system. Two versions of PINS system were used; one excited the test object with neutrons from a 252C f source and one with neutrons from a portable deuterium-deuterium (DD) neutron generator. The sets of measurements and simulations with the two different neutron sources provided experimental data for development of automatic fill identification algorithms as well as providing a good test of the agreement between the MCNP calculations and experiment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

A Cyclic-Plasticity-Based Mechanistic Approach for Fatigue Evaluation of 316 Stainless Steel Under Arbitrary Loading

In this paper, a cyclic-plasticity based fully mechanistic fatigue modeling approach is presented. This is based on time-dependent stress-strain evolution of the material over the entire fatigue life rather than just based on the end of live information typically used for empirical S~N curve based fatigue evaluation approaches. Previously we presented constant amplitude fatigue test based related material models for 316 SS base, 508 LAS base and 316 SS- 316 SS weld which are used in nuclear reactor components such as pressure vessels, nozzles, and surge line pipes. However, we found that constant amplitude fatigue data based models have limitation in capturing the stress-strain evolution under arbitrary fatigue loading. To address the above mentioned limitation, in this paper, we present a more advanced approach that can be used for modeling the cyclic stress-strain evolution and fatigue life not only under constant amplitude but also under any arbitrary (random/variable) fatigue loading. The related material model and analytical model results are presented for 316 SS base metal. Two methodologies (either based on time/cycle or based on accumulated plastic strain energy) to track the material parameters at a given time/cycle are discussed and associated analytical model results are presented. From the material model and analytical cyclic plasticity model results, it is found that the proposed cyclic plasticity model can predict all the important stages of material behavior during the entire fatigue life of the specimens with more than 90% accuracy

316 Stainless Steel↗
DOE OSTI2022

MECHANICAL PERFORMANCE EVALUATION OF THE PRINTED CIRCUIT HEAT EXCHANGER CORE EXPERIMENTS UNDER TENSION AND PRESSURE LOADING

The printed circuit heat exchanger (PCHE) has small channels with high surface area, making them an efficient solution for next-generation nuclear plants (NGNPs). These PCHEs are fabricated through a diffusion bonding process. This fabrication step changes the microstructure of wrought metal plates. The current ASME design code does not support the PCHE design for NGNPs due to a lack of test data. Hence, there has been initiative towards elevated temperature mechanical property characterization of the diffusion bonded material. One of the most common channel shapes is a semicircular channel with sharp corners. These corners act as a stress riser at the diffusion bonding interface. Evaluating elevated temperature mechanical performance of diffusion bonded material in the presence of stress risers is an essential step towards the ASME code development of PCHE design. This study selected two specimen geometries: the first is a PCHE bar specimen for tensile loading with three rows and three columns of channels, and the second is a lab-scaled PCHE with six rows and eight columns of channels. A set of elevated temperature monotonic and cyclic tests were conducted on the PCHE bar specimen to evaluate the mechanical performance under axial tensile loadings to study the failure mechanism. The lab-scaled PCHE specimens were tested under overpressure loads at room temperature, and pressure creep and pressure creep-fatigue loadings to mimic the realistic loading conditions observed in typical NGNPs. The X-ray scans of channeled specimens show interesting observations. The test results and observations are presented in the paper.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Neutron Leakage Spectrum from a Pulsed Uranium Sphere [Slides]

The neutron leakage spectrum from a pulsed uranium sphere was calculated with MCNP5 and found to be in good agreement with previous “MCN” simulations and experimental data. This example is proposed as another time-of-flight (TOF) test problem for NUEXS, a code for diagnostic simulation maintained by LANL Group P-23.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

AGR-5/6/7 Experiment Monitoring and Simulation Progress

Advanced Gas Reactor (AGR)--5/6/7 is the last of a series of AGR experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tri-structural isotropic (TRISO) low-enriched fuel for use in high temperature gas-cooled reactors. The test train contains five separate capsules that are independently controlled and monitored. Each capsule contains multiple 12.51 mm long compacts filled with low-enriched uranium carbide/oxide (UCO) TRISO fuel particles. The objectives of the AGR-5/6/7 experiment are to: 1. Irradiate reference-design fuel particles to support fuel qualification. 2. Establish operating margins for the fuel beyond normal operating conditions. 3. Provide irradiated fuel performance data and irradiated fuel samples for post-irradiation examination (PIE) and safety testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Select Transient Metallic Fuel Bison Benchmarks

To explore the current capabilities of Bison with respect to transient metallic fuel modeling, four transient experiments are compared with their respective Bison model. The four experiments are the OPT-1 experiment of the Experimental Breeder Reactor II~(EBR-II) and the M5, M6, and M7 experiments of the Transient Reactor Test facility (TREAT). Each benchmark is a 2D-RZ geometry of a single pin employing coupled multiphysics. Bison models of the benchmark agree well and illustrate areas needing development. The areas of suggested development deal with better material models, better coolant channel module for transients, proper treatment of material melting, improved feedback with failure models, and an ability to set initial conditions from post irradiation examination (PIE) data. Two Fast Flux Test Facility (FFTF) pins are also modeled with M7 transient conditions to illustrate Bison being applied to investigate potential experiments.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Development of Gamma Background Radiation Digital Twin with Machine Learning Algorithms: Application of Unsupervised Machine Learning to Detection of Anomalies and Nuisances in Gamma Background Radiation Environmental Screening Data

Environmental screening of gamma radiation consists of detecting weak nuisance and anomaly signal in the presence of strong and highly varying background. In a typical scenario, a mobile detector-spectrometer continuously measures gamma radiation spectra in short, e.g., one-second, signal acquisition intervals. The measurement data is a 2D matrix, where one dimension is gamma ray energy, and the other dimension is the number of measurements or total time. In principle, gamma radiation sources can be detected and identified from the measured data by their unique spectral lines. Detecting sources from data measured in a search scenario is difficult due to the highly varying background because of naturally occurring radioactive material (NORM), and low signal-to-noise ratio (S/N) of spectral signal measured during one-second acquisition intervals. The objective of this work is to explore unsupervised machine learning (ML) algorithms for development of a digital twin of gamma radiation background, and for detection and identification of weak nuisances and anomalies events in the presence of highly fluctuating background. In one segment of work, we developed a gamma background estimation model using a Longshort term memory (LSTM) network for one-step CPS time series prediction. The LSTM model was validated with two data sets of measurements from two independent NaI detectors positioned on a mobile platform. The data sets contained background radiation only and no orphan isotope sources. The LSTM model was constructed and tested using data from one of the detectors. Performance of the LSTM model was validate through one-step prediction of CPS time series of another NaI detector without re-training. This approach allows to create a digital twin for nuclear background estimation. Using LSTM, it could be possible to detect a source through subtraction of the estimated counts from the measured background. In another segment of work, we investigated detection of gamma emitting sources in the presence of complex background using unsupervised machine learning. Spectral lines of isotopes are difficult to observe in one-second measurements. Averaging over the entire measurement campaign data set reveals spectral lines of most common background isotopes. Spectral lines of orphan sources, which might appear only in a few measurements during the campaign, will be washed out if averaging is performed over the entire measurement data set. The approach we have explored consists of extracting one-second measurements containing weak spectral features through data clustering. Averaging one-second spectra in a cluster should reveal the presence of anomaly sources. We created two ML models using K-means clustering and Neural Network Self-organizing Map (SOM). Performance of these ML models was benchmarked using search data. One data set contained 137 Cs source, and another dataset contained 131 I source.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Large Surface Explosion Coupling Experiment - SNL Remote Optical.

Two surface chemical explosive tests were observed for the Large Surface Explosion Coupling Experiment (LSECE) at the Nevada National Security Site in October 2020. The tests consisted of two one-ton explosions, one occurring before dawn and one occurring mid- afternoon. LSECE was performed in the same location as previous underground tests and aimed to explore the relationship between surface and underground explosions in support of global nonproliferation efforts. Several pieces of remote sensing equipment were deployed from a trailer 2.02 km from ground zero including high-speed cameras, radiometers and a spectrometer. The data collected from these tests will increase the knowledge of large surface chemical explosive signatures.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Our teams design high-resolution MeV x-ray radiography systems for National security missions [Poster]

The NNSS has been at the forefront of custom high-fidelity x-ray/gamma radiographic imaging solutions that serve our national security for over five decades. Our radiography team utilizes expertise in physics modeling and analysis, along with optical, mechanical, and electrical design, in close collaboration with our customers, to develop imaging methods and capabilities that go beyond their needs. Conceptual designs are developed through R&D efforts to provide solutions for the specific problem at hand. Field systems are designed and built in-house, then qualified utilizing a range of facilities across the National Security Enterprise. Radiographic imaging systems are deployed by our team in the most challenging environments. The NNSS has a strong math and programming team that provides novel on-site image analysis methods that extract crucial information from data returned in field tests.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Hanford Low-Activity Waste Glass Minor Component Concentration Boundary Expansion

This study investigated the effects of increasing concentrations of minor components on simulated Hanford low-activity waste (LAW) glass properties to allow for LAW glass composition envelope expansion. A test matrix of ten glasses was generated by spiking three previously studied simulated Hanford LAW glass compositions with increasing concentrations of minor component mixes up to a total concentration of 2 wt%. In this study, the ten glasses were fabricated, and key properties were measured including crystal formation after canister centerline cooling (CCC), crystallinity as a function of temperature, density, viscosity, electrical conductivity, glass durability using the Vapor Hydration Test (VHT), product consistency using the Product Consistency Test (PCT), and sulfur solubility. The data from these tests are presented in this report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗