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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 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

Non-invasive Pipe Pressure Monitoring for Safeguards

Our project explores a new solution to complement current safeguards toolbox by independently, and noninvasively, determining the pressure inside uranium hexafluoride (UF 6 )-carrying Gas Centrifuge Enrichment Plant (GCEP) pipes. Proof-of-principle, non-invasive pipe pressure monitoring tests were conducted through measurements of hoop and axial strain using an optical-interference dilatometry technique based on fiber Bragg gratings (FBGs) inscribed in the core of telecommunication-type 125 µm diameter optical fibers. The fibers were affixed to commercial stainless steel and aluminum tubes and pipes of the type expected to be found at GCEP’s, each measuring four inches in diameter. When the internal air pressure in the pipes changes their external diameter and length follow, and the pitch in the FBG shrinks and/or expands accordingly. In turn, these changes result in a shift of the light wavelength reflected by the FBGs. Several FBGs physically attached to the pipe, as well as control FBGs not attached, were interrogated during the process using state-of-the-art commercial interrogators that operate on swept-wavelength laser technology. Strains measured were contrasted against calculations based on the known elastic and mechanical properties of the pipes. Good agreement is found, and strategies are proposed for further improvement of the experimental resolution as well as future implementation of this novel approach for GCEP safeguards.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Assessment of Self-interrogation Safeguards Signatures for Pebble Bed Reactor Fuel

New reactor designs, such as pebble-bed reactors, present challenges for the safeguarding of fissile material. Due to the multi-pass fuel circulation design and the non-uniform path a pebble may traverse through a pebble-bed reactor, there will be variations in the irradiation history of spent fuel pebbles. The standard approach of estimating actinide quantities based on fission gamma spectra or neutron emission paired with depletion modeling may yield uncertainties too large for safeguards and material accountancy. This project investigated, through modeling and experimentation, the potential of neutron self-interrogation of spent fuel pebbles as an innovative method to implement materials accountability. As an example, our feasibility studies indicate that the mass of U-235, U-238, Pu-239 and Pu-241 can be predicted to 4.1, 0.86, 13 and 13 % accuracy, respectively, when measuring 100 closely packed end-of-life spent fuel pebbles over approximately 12 days using a 4π counting geometry.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

An Assessment of the Safeguards Implications of Particle and Alloy Formation in MSRs

Molten salt reactors (MSRs) present unique challenges for applying safeguards measures and differ significantly from traditional nuclear reactors. Although the special fissionable materials within MSRs principally exist as dissolved species within the fuel salt, key actinides of interest may also be converted to solid phases including alloys and precipitated oxides via chemical and electrochemical reactions. These solid-phase components will complicate the nuclear material accounting and control (NMAC) approaches that have, so far, primarily targeted measurements of the liquid salt itself. For example, most of the proposed monitoring and characterization methods for MSRs are reliant on homogeneity of the salt and the ability to maintain accountable materials in solution. Solid deposits and particulates could invalidate the assumption of homogeneity and necessitate the use of new models and measurement tools to permit a full accounting of all nuclear material in an MSR. To address this, the likelihood of solids formation in MSRs, estimated the rate at which these solids could form, and assessed the monitoring tools that are available to account for the presence of these materials were examined. Specifically, three solid forming mechanisms were investigated, including (1) the accumulation of actinides within the MSR’s structural materials due to alloying, (2) the accumulation of actinides within noble metal fission product particulates, and (3) the generation of actinide oxides from interactions with salt impurities. These modeling-informed studies were completed with considerations to both fluoride- and chloride-containing salts. It was determined that, although measurable amounts of special fissionable material were expected to alloy with the structural elements and fission products, the predicted amounts were below levels considered as significant quantities (SQ) by the International Atomic Energy Agency (IAEA). However, it was determined that sufficiently large oxygen concentrations could potentially lead to significant quantities of special fissionable material being converted to solid oxide phases. As such, a comparative study of technologies that are commercially available to assist in monitoring and accountancy of this material was presented. Policy implications, design recommendations, and suggestions for follow-on experimental work were also presented to further assist reactor vendors in designing and controlling the fuel salt chemistry to address potential safeguards considerations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Plutonium Signatures in Molten-Salt Reactor Off-Gas Tank and Safeguards Considerations

Fluid-fueled molten-salt reactors (MSRs) are actively being developed by several companies, with plans to deploy them internationally. The current IAEA inspection tools are largely incompatible with the unique design features of liquid fuel MSRs (e.g., the complex fuel chemistry, circulating fuel inventory, bulk accountancy, and high radiation environment). For these reasons, safeguards for MSRs are seen as challenging and require the development of new techniques. This paper proposes one such technique through the observation of the reactor’s off-gas. Any reactor design using low-enriched uranium will build up plutonium as the fuel undergoes burnup. Plutonium has different fission product yields than uranium. Therefore, a shift in fission product production is expected with fuel evolution. The passive removal of certain gaseous fission products to the off-gas tank of an MSR provides a valuable opportunity for analysis without significant modifications to the design of the system. Uniquely, due to the gaseous nature of the isotopes, beta particle emissions are available for observation. The ratios of these fission product isotopes can, thus, be traced back to the relative amount and types of fissile isotopes in the core. This proposed technique represents an effective safeguards tool for bulk accountancy which, while avoiding being onerous, could be used in concert with other techniques to meet the IAEA’s timeliness goals for the detection of a diversion.

Dunkle, Nicholas (ORCID:000000027323011X)↗
NASA NTRS1984

NASA guidelines for assuring the adequacy and appropriateness of security safeguards in sensitive applications

The Office of Management and Budget (OMB) Circular A-71, transmittal Memorandum No. 1, requires that each agency establish a management control process to assure that appropriate administrative, physical and technical safeguards are incorporated into all new computer applications. In addition to security specifications, the management control process should assure that the safeguards are adequate for the application. The security activities that should be integral to the system development process are examined. The software quality assurance process to assure that adequate and appropriate controls are incorporated into sensitive applications is also examined. Security for software packages is also discussed.

Tompkins, F. G.↗
NASA NTRS2004

Security Management and Safeguards Office

The Security Management and Safeguards Office at NASA is here to keep the people working in a safe environment. They also are here to protect the buildings and documents from sabotage, espionage, and theft. During the summer of 2004, I worked with Richard Soppet in Physical Security. While I was working here I helped out with updating the map that we currently use at NASA Glenn Research Center, attended meetings for homeland security, worked with the security guards and the locksmith. The meetings that I attended for homeland security talked about how to protect ourselves before something happened, they told us to always be on the guard and look for anything suspicious, and the different ways that terrorist groups operate. When I was with the security guards I was taught how to check someone into the base, showed how to use a radar gun, observed a security guard make a traffic stop for training and was with them while they patrolled NASA Glenn Research Center to make sure things were running smooth and no one was in danger. When I was with the lock smith I was taught how to make keys and locks for the employees here at NASA. The lock smith also showed me that he had inventory cabinets of files that show how many keys were out to people and who currently has access to the rooms that they keys were made for. I also helped out the open house at NASA Glenn Research Center. I helped out by showing the Army Reserves, and Brook Park's SWAT team where all the main events were going to take place a week before the open house was going to begin. Then during the open house I helped out by making sure people had there IDS, checked through there bags, and handed out a map to them that showed where the different activities were going to take place. So the main job here at NASA Glenn Research Center for the Security Management and Safeguards Office is to make sure that nothing is stolen, sabotaged, and espionaged. Also most importantly make sure all the employees here at NASA are in a safe environment.

Bewley, Nathaniel M.↗
DOE OSTI2022

A Tool Kit for Generating Simulated Radiation Measurements for Advanced Reactor Safeguards and Security

A tool kit was developed to simulate and analyze passive radiation measurements of molten salt reactor (MSR) operations to support development of nuclear safeguards approaches for this emerging reactor technology. A Transient Simulation Framework of Reconfigurable Modules (TRANSFORM) multiphysics simulation of an MSR produces time-dependent isotopic inventories at user-selected locations within the model. The tool kit implements the Gamma Detector Response and Analysis Software (GADRAS) application programming interface to inject the TRANSFORM isotopic inventories extracted/processed by a Python pipeline into GADRAS models of user-defined geometries. The TRANSFORM inventories are the source terms used to obtain synthetic measurements from GADRAS-defined detectors. The speed of TRANSFORM and GADRAS simulations enables surveying the large design space of MSRs (e.g., fuel type, fuel salt composition, number of loops) and the plethora of measurements (e.g., location, detector type, and collimation) within the reactor. This has enabled timely assessment of the various measurement locations and detectors to identify the most effective and efficient safeguards approach for a specific MSR design. Lastly, the tool kit also simulates extracted samples that can be aged to a desired dose, enabling stakeholders to optimize a measurement plan to use sample analysis as an element within a broader material accountancy plan.

Westphal, Greg↗
DOE OSTI2020

Leveraging Machine Learning Capabilities for the Characterization of Irradiated Uranium: A Case Study of Analysis Methods for Nuclear Safeguards and Nuclear Forensics

Nondestructively determining the initial enrichment of irradiated uranium is a complex and laborious multivariable problem due to the presence of fission products. This work demonstrates the capabilities of machine learning to analyze gamma-ray spectral data to determine initial enrichment without knowledge of the decay time of the sample. The approach developed is agnostic to the particular scenario and is applicable to a wide variety of applications in nuclear forensics and nuclear safeguards. We irradiated 5 mg uranium standard reference materials at discrete enrichment values ranging from 0.02% to 97% 235 U (weight percent) in UT Austin’s Nuclear Engineering Teaching Laboratory TRIGA Mark II 1.1 MW research reactor, allowed each to decay for 8 hours, and then measured each sample via gamma-ray spectrometry for 50 hours post-irradiation yielding 1,400 individual gamma-ray spectra discretized into 8,192 energy bins. We then trained decision trees models to analyze individual gamma-ray spectra and estimate the associated initial enrichment without knowledge of the time since end of irradiation. We evaluated the performance of the models with a reserved test set not used for training or calibrating the model. A decision tree model constructed with this procedure achieved a mean absolute error in initial enrichment determination of 2.3% (weight percent 235 U). Next, we implemented a principal component analysis pre-processing routine of the gamma-ray spectrometry data to reduce the dimensionality of the dataset from 8,192 channels in the spectrum to 10 principal components while retaining over 99% of the inherent variance in the data. Decision tree models constructed with these data demonstrated decreased mean absolute error in enrichment determination, reduced computational time, and decreased complexity. A single decision tree model constructed with this procedure achieved a mean absolute error in initial enrichment determination of 0.05% (weight percent 235 U). Furthermore, we analyzed these models with learning curves to ensure that overfitting did not occur. The capabilities provided by these models can be naturally extended to other application-focused measurements in the fields of nuclear safeguards, nuclear forensics, and nuclear non-proliferation.

Drescher, Adam↗
DOE OSTI2020

A highly-enriched 244 Pu reference material for nuclear safeguards and nuclear forensics measurements

A highly-enriched 244 Pu isotope dilution reference material has been prepared and characterized for metrologically traceable measurements of very small quantities of plutonium. The amount of plutonium in samples associated with nuclear safeguards and nuclear forensic measurements can be significantly less than 1 ng. Accordingly, the ability to quantify the amount and isotopic composition of plutonium from a single mass-spectrometric analysis is particularly desirable. The highly-enriched 244 Pu reference material, described here, will minimize the magnitude of spike corrections necessary to obtain accurate information on plutonium isotopic composition from isotope dilution measurements.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

High bias machine learning for antineutrino-based safeguards for small reactors

The statistical methods used for antineutrino detection will need to be improved to effectively monitor the inventory of next-generation nuclear reactors. In this sensitivity study, we evaluate machine learning models compared to previously used statistical approaches to identify diversion scenarios in a simulated Advanced Fast Reactor (AFR)-100. A chi-square goodness-of-fit technique, which individually compares the simulated antineutrino yields to the expected antineutrino yield, resulted in precise but low diversion detection probability. Various support vector machine (SVM) models were applied with diverse training datasets to evaluate the robustness of the method towards unexpected or “unseen” diversion scenarios. Furthermore, our results indicate that while the SVM models significantly improved the detection probability of near-field antineutrino-based safeguards, up to a probability of ~0.04, for the simulated small reactor, the detection system still needs improvements to reach the 0.2 detection limit established by the International Atomic Energy Agency.

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

Validation and demonstration of the AEFC as a practical safeguards tool for inventory verification

The Advanced Experimental Fuel Counter (AEFC) is a nondestructive assay (NDA) instrument designed to determine the residual fissile mass in irradiated fuel assemblies for safeguards verification purposes. This is done by actively interrogating an assembly with a neutron source and measuring the total (Singles) and correlated (Doubles) neutron count rates resulting from induced fissions in the irradiated nuclear fuel and relating those rates to the residual fissile mass using calibration curves. Comprehensive NDA measurements of the irradiated fuel inventory at Israeli Research Reactor 1 (IRR-1) were taken with the AEFC to validate a set of previously developed calibration curves. During the campaign, measurements were acquired of 32 standard fuel assemblies and three control assemblies in just nine days. This is a significant majority of the research reactor's irradiated fuel inventory and the largest data set gathered by the AEFC to date. Many of the fuel assemblies measured during the campaign had much shorter cooling times than those assemblies previously measured with the instrument. Calibration curves developed from previous AEFC deployments were used to determine the residual 235 U mass in the measured standard fuel assemblies. The results of the campaign demonstrated that the AEFC can be used to estimate the 235 U mass remaining in a large number of irradiated fuel assemblies with 1%–5% uncertainty in a reasonable amount of time despite operating in a high dose environment.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2022

Feasibility Study of Millimeter Wave Radars For Safeguards Applications

Containment and surveillance are fundamental measures in nuclear safeguards. Techniques such as video surveillance and laser curtain for containment provide effective monitoring in areas where maintaining continuity of knowledge is required. These systems, however, can be susceptible to loss of monitoring capabilities under certain environmental conditions such as poor visibility (i.e. low light conditions, smoke, fog, etc.) or extended power loss past the duration that the backup power system is designed for. Brookhaven National Laboratory has been investigating the feasibility of millimeter waves (mmWave) as a new perimeter seal in which radio frequency waves in the range of 60-64 GHz are used to detect and monitor objects of interest. Signals in this frequency range are not susceptible to environmental conditions. For proof-of-concept tests, mmWave sensors from Texas Instruments (TI), specificallyIWR6843, are used in a test bed at BNL's Waste Management facility to simulate the operations at nuclear facilities. The unique design of TI mmWave sensors requires less memory and power consumption compared to counterpart systems. These devices are capable of exporting 3D point-cloud data, which is visualized graphically and compared to videos recorded at the same time to validate the performance of the mmWave sensor. A set of experiments were planned to test the feasibility of the mmWave in this application, including monitoring static containers in a storage area and detecting intrusions at the boundaries of the area. In addition, the experiments also identify potential blind spots relative to sensor position and utilize multiple operating sensors simultaneously to reduce or eliminate such blind spots. The optimal positioning of multiple sensors was determined for the experimental room configuration. In this paper, we will discuss the details of this novel perimeter sealing concept and present the test results.

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

Development of spectroscopy apparatus for molten salt reactor safeguards and material characterization

A molten salt reactor (MSR) produces nuclear fission energy using mixture of fissile and non-fissile salts. Salt in the reactor functions as the nuclear material and the coolant allowing for a safer design. International nuclear safeguards is the monitoring of nuclear reactors, nuclear fuel, and irradiated material to prevent the proliferation of nuclear weapon grade material in hostile foreign nations. The mixture of salt creates a proliferation risk as it generates a conglomeration of fuel, daughter nuclei, and corrosive material. A method to monitor and measure the materials present in the reactor is essential to the scalable deployment of MSRs. The objective of this project was to design a spectroscopy apparatus for aqueous salt so it can be measured and collected consistently in the aerosol phase.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2026

Conceptual design study of neutron detectors for safeguards measurement of an irradiated pebble

Nuclear material control and accounting (MC&A) of pebble-bed reactors (PBRs) is challenging because a PBR utilizes hundreds of thousands of identical, unmarked pebbles that are continuously recirculated through the core. To develop tools that enable the implementation of international safeguards, especially in the context of MC&A of spent pebbles, we designed and simulated three neutron detection concepts to determine fissile content in individual pebbles: a differential die-away (DDA) detector, a californium interrogation prompt neutron (CIPN) detector, and a passive neutron albedo reactivity (PNAR) detector using Monte Carlo calculations. Burnup calculations were performed on the spent pebbles from the PBMR-400 classic PBR. The varying neutron and gamma source terms, and isotopic compositions in the spent pebbles calculated at various burnup levels were used in the neutron detector models. DDA was found to be sensitive to the number of passes a pebble has had through the core and to the fissile content contained in a spent pebble. Optimization in the DDA design further increased the neutron count rates and thus reduced counting uncertainty. Meanwhile, passive neutron counting using the same detector body could distinguish pebbles with different numbers of passes, but its response was dominated by neutron-emitting actinides and was not sensitive to fissile content. On the other hand, the PNAR technique was not viable for a single pebble but performed reasonably for a 27-pebble array, which suggested potential use for verification measurements of containers filled with 27 or more spent pebbles.

CIPN↗
DOE OSTI2022

Improved evaluation of safeguards parameters from spent fuel measurements with the Differential Die-Away (DDA) instrument

The Differential Die-Away (DDA) technique is a highly sensitive non-destructive assay method for characterizing and detecting the presence of fissile material within an item of interest. DDA utilizes a series of pulses from a neutron generator (NG) to actively interrogate an item of interest. The die-away time of the neutron population induced by this active interrogation and the integral of the total differential die-away signal can be used to characterize items such as nuclear waste drums and spent nuclear fuel assemblies. In this work, Los Alamos National Laboratory (LANL) conceptualized, designed, and fabricated a DDA instrument that was deployed for field test measurements at the Central Interim Storage Facility for Spent Nuclear Fuel (Clab) in Oskarshamn, Sweden. The instrument performed multiple static measurements at fixed locations and dynamic axial scans of 15 pressurized water reactor (PWR) and 10 boiling water reactor (BWR) spent fuel assemblies, collecting both passive and active measurement data. The static assays of the assemblies measured the differential die-away signal, die-away time, and total passive neutron emission rate to create calibration curves for the evaluation of assembly multiplication, burnup, initial enrichment, effective fissile mass, and total elemental plutonium mass. Each calibration curve was optimized by minimizing the relative root mean square error (RRMSE) of assembly assay results compared to declared assembly parameters. The same quantities were also measured with the axial scans, and the resulting data were applied in two ways: (1) in the creation of calibration curves to improve evaluation of the same safeguards parameters as static assays, and (2) for comparison to simulation. In most cases, across both PWR and BWR assemblies, axial scan data improved the estimation of the above parameters, quantified by decreasing the calibration curve RRMSE. These axial scan results demonstrate the ability of the DDA instrument and analysis method to characterize spent PWR and BWR fuel as well as, or better than, a static assay of the same assembly. Furthermore, the DDA instrument’s unique ability to obtain both active and passive data in a single, axial scan of an entire spent fuel assembly represents a more efficient and accurate way of assaying spent fuel for verification purposes. These results represent a significant advancement for characterizing spent nuclear fuel compared to current technologies.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Neutron resonance transmission analysis prototype system for thorium fuel cycle safeguards

Emerging thorium-based reactor designs and fuel cycles present challenges to traditional non-destructive assay techniques used in international safeguards. Specifically, assaying the masses of 233 U and 235 U when they are present together in samples with high gamma ray backgrounds is difficult because of similar passive neutron signatures and relatively weak gamma-ray emissions of 233 U. The Pacific Northwest National Laboratory (PNNL) and the Massachusetts Institute of Technology (MIT) are developing a compact neutron resonance transmission analysis (NRTA) system as one potential solution to these challenges. The NRTA technique provides isotopic information for a sample via neutron time-of-flight (TOF) measurements that exploit a sample’s epithermal neutron resonance cross-sections. A recently developed portable NRTA system uses a commercially available, pulsed deuterium-tritium neutron generator with a ~2 m flight path and a GS20 lithium glass scintillator detector. Finally, this paper describes the prototype NRTA system design, a refined radiation transport model of the system, preliminary measurements with thorium and uranium sources, and demonstration of a quantitative isotopic estimation algorithm.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗