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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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636 records · Page 36
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Decontamination Experiments and Simulations Supporting Recovery After Widespread Contamination - 18685

Clean-up after an uncontrolled nuclear material release differs from common decontamination practices in the nuclear industry and military combat scenarios. Uncontrolled releases, particularly acts of terror, require specific protective actions guides (PAGs) to optimize short and long-term repercussions. Ideally, non-destructive decontamination methods would be employed. To help fulfill this need, the IWATERS (Integrated Wash Aid Treatment Emergency Reuse System) was proposed in 2014 for remediation outside the immediate blast area. 'Wash-Aid' is hosed onto affected areas, and the used wash solution is collected and recycled through sand/clay beds. To help characterize the IWATERS technology, the below was completed: - Decontamination experiments with a pressurized wash solution using Cs-137, Sr- 85, and Eu-152 contaminants on porous urban materials. - Sensitivity analyses on recycling bed variables using GoldSim{sup TM} CT. - A proposed framework that expands the recycling bed model to include probabilistic simulation of immediate recovery efforts. Removal efficacy of tap water and 0.5 M KCl were compared for Cs, Sr, and Eu across asphalt, brick, and concrete. The 0.5 M KCl solution significantly outperformed tap water in every case except for Eu on concrete and Cs on brick. Trends in the high-pressure decontamination with salt additive compared to low pressure decontamination show evidence of physical and chemical removal. Variations in cleaning rate yielded no significant trend. Pressure changes (by adjusting nozzle-to-coupon distance) were only significant for the highest pressure. Hypothetical Contamination Scenario: To tie the decontamination experiments and recycling bed model together, a small contamination scenario was created. Imagine the sides of a small concrete building uniformly contaminated with Cs-137. Power washing the building with 0.5 M KCl solution would result in 857 L of contaminated solution generated in 185 minutes. Two sand clay beds were designed to clean the used solution. With a sand mass fraction of 0.9, a 10 ft{sup 2} bed 4 ft tall could recycle the solution in 1.05 hrs. Changing the sand mass fraction to 0.5 yielded a 2.5 ft{sup 2} bed 4 ft tall (∼55-gallon drum) that recycled the solution in 7.43 hrs. This exercise showed how the sand/clay bed design can easily change to match the primary remediation goals. The second sand/clay bed would reduce the waste generation while the first bed reduces the remediation time. Conclusions: Variations on speed and nozzle distance (pressure) yielded insignificant differences in efficacy except for the closest distance. High-pressure washing showed increased decontamination efficacy with KCl additive in the majority of cases. Except for bed composition, key recycling bed parameters can be varied on-the-fly. This allows for optimization of, for example, time and waste generation. The hypothetical contamination scenario showed pairing decontamination experiment results with modeling tools can lead to increased preparedness for a large-scale contamination event. With collaboration and time, an expanded recovery/remediation model could aid in development of effective protective action guides.

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

Highlights from the Postirradiation Examination of AFC-3A and AFC-3B

Highlights from the postirradiation examination of the AFC-3A and AFC-3B irradiation test are presented. Baseline non-destructive and destructive examinations of these tests have been completed at this time. Results from visual examination, neutron radiography, gamma spectrometry, dimensional inspection, fission gas release, chemical burnup analysis, and optical microscopy have been collected. All the rodlets from these irradiations appear to have performed well during irradiation. Baseline PIE has been completed on rodlets from AFC- 3A, and AFC-3B. This includes visual exams, neutron radiography, gamma spectrometry. dimensional inspection, fission gas release evaluation, optical microscopy. microhardness evaluation. and chemical analysis. The results compiled here are a sampling of the full results. Baseline PIE provides an engineering or macroscopic scale evaluation of fuel performance and prepares samples for a more detailed phenomenological study of observed performance. Advanced PIE techniques are desired to obtain microscopic elemental and structural information both in the fuel, in the cladding, and in the interface where the composition of the FCCI layer is of great interest. The AFC-3 tests do appear to have performed roughly as designed during the irradiation. The FCCI behavior of Mo alloy metallic fuel versus Zr alloy metallic fuel has been confirmed. Results from this irradiation such as the FCCI observed in AFC-3A R1 and AFC-3A R2 can be used to eliminate Mo alloys as a choice in future irradiations where ultra-high burnup is a stated goal. Annular fuels are a preferred method for many high burnup core designs. The redistribution behavior of the annular U-Zr alloys should be studied further to better understand how this might impact high burnup annular fuel performance. The thermal bonding between annular fuel and cladding should also be studied further as there appear to be localized areas of FCCI forming in both AFC-3A R4 and AFC-3B R2, while no FCCI was observed in the Zr-based Na-bonded or Mo-based low smear density Na-bonded fuel. The performance of the low smear density. solid fuels is less clear. The neutron radiography and optical microscopy cross sections suggest the fuel is not forming a radially symmetric structure which may be causing undesirable localized temperature variations. Further study both by electron microscopy and by simulation to evaluate the implications of the non-symmetric fuel form are needed to better understand whether this fuel form merits further study under irradiation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Characterization of Fresh Nuclear Fuel using Time-Of-Flight Neutrons

Neutrons offer excellent characterization capabilities for nuclear fuels and can provide non-destructive insight on micro structure, chemistry and crystallographic phase. Measurements are possible at reactors or pulsed sources, but the inherent energy resolution implicit at pulsed sources offers unique opportunities. Recent advances in 2D neutron detection enabled new capabilities of material characterization using neutron imaging and tomography for nuclear materials. Moreover, time-of-flight (TOF) neutron measurements can enable isotope specific tomography by deriving isotope densities from transmission spectra exhibiting neutron resonances, which are recorded for each pixel. Tomographic algorithms allow 3D reconstructions of the spatial distribution of isotopes, density and features in typical fuel pellet samples. The ability to select radiographs for different energies allows characterizing samples which are opaque for thermal neutrons, such as enriched fuels or fuels containing plutonium, by selecting epi-thermal neutron energies and by knowledge of the total neutron cross sections of the constituents, isotopic enrichment levels can be determined. Nuclear cross sections are typically not calculated directly and databases such as the Evaluated Nuclear Data File (ENDF) database provide this information with cross sections determined by standardized experimental measurements, such as neutron transmission measurements for individual isotopes. Utilizing the information in the ENDF database, the 'inverse' problem can be solved by assuming cross section parameters to be known parameters and areal densities can be determined. This work presents recent advances at the Los Alamos Neutron Science Center (LANSCE) facility at Los Alamos National Laboratory (LANL), which provides the infrastructure to handle radioactive samples, in the development of fuel characterization with TOF neutrons. Examples include recent work on prototype accident tolerant nuclear fuels (ATF). The potential application of these techniques for nondestructive examination of irradiated fuel is under consideration. Non-destructive access to materials features for specimens embedded in containers, particularly for hazardous samples, such as a radioactive nuclear fuel, is indispensable to characterize and ultimately license new fuel forms. Energy-resolve neutron imaging is a new tool for this research field. The results reported here present a new imaging technique, setting a milestone in the capabilities of measuring non-destructively absolute quantities of physical properties. The enabling physics behind this new technique are nuclear resonances. We have demonstrated how absolute areal densities can be extracted from radiographs utilizing pulsed neutron sources. This technique provides information that is otherwise difficult and costly to obtain. In the case of nuclear fuels, the results from energy-dispersive neutron imaging may guide destructive examination of irradiated fuel. This summery does not describe recent analysis, which is ongoing. Particularly, the implementation of calibration standards has improved the determination of enrichment levels without additional calibration. The analysis of the samples was performed with full tomographic reconstructions for isotope specific data and presents a first approach in absolute isotope specific 3D reconstruction. This work was predominantly funded by and received substantial support from the Fuel Cycle R and D program at Los Alamos National Laboratory. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Integrating Predictions for Improving Defect Classification Accuracy in NDT-based Assessment of Concrete - 20229

There is an increasing need to create predictive models for defect classification in concrete using the output of non-destructive testing (NDT) techniques. Recent advancement of machine-learning algorithms has offered several techniques for developing classification models for different types of data sets. However, the performance of these algorithms is very uncertain, mainly when applied to small and noisy datasets. For example, when human access is limited (e.g., nuclear facility), robot-based NDT is preferred. But compared to manual tests with humans present on site, the data sets are small, and more noise can exist. Therefore, it is imperative to develop new approaches to ensure a consistently high classification accuracy for inadequate data sets. This study explores the classification performance on NDT dataset using classifiers from different machine-learning algorithms, namely k-Nearest Neighbor (kNN), Decision Tree, Naive Bayes, Logistic Regression, and Support Vector Machine (SVM). The authors further integrated the predictions from these classifiers using proposed methods. The integration strategy combines the output of the classifiers based on two different measures, accuracy, and performance (ACC and PERF), using equations such as sum, average, and square-root-of-sums-of-squares (SRSS). Our results reveal varying classification accuracies across individual classifiers with different misclassifications across the test data set. The integration strategy provided significant improvement in the classification accuracy compared to the individual classifiers. Furthermore, the results indicate minimal variation across the integration methods as compared to the variation across the individual classifiers. To conclude, prediction integration offers a unique approach for combining the output of multiple classifiers to create redundancies with the potential of achieving high classification performance and improved reliability in predictive models for defect detection in concrete. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Post-Irradiation Examination for Conventional and Advanced Nuclear Fuel

During irradiation in a nuclear reactor, the fuel pellet experiences non-uniform fission rate and strong thermomechanical loads, which result in alterations and restructuring occurring non-homogeneously in the fuel volume. The irradiated fuel is characterized by a highly heterogeneous structure, with a correspondingly heterogeneous distribution of phases and properties. It is a big technical challenge for a hot cell facility to be able to cover effectively a broad variety of fuel types and compounds, characterized by different compositions, physico-chemical properties, geometries and configurations. The first challenge is to provide testing environments and procedures which do not alter or degrade the original irradiated fuel properties and are technically suitable for hot cell or shielded facility operation. For instance, non-oxide fuel compounds require protective atmosphere to avoid sample oxidation during storage, preparation and testing. Moreover, suitable adaptations of tools and measuring devices must be available to characterize fuel configurations which differ from the conventional cylindrical pellet-cladding geometries (e.g. spherical fuel elements for pebble bed HTR and TRISO coated fuel particles). Finally, adequate scientific methods of investigation must be available to measure properties and study the behaviour of highly heterogeneous irradiated fuel. In addition to basic techniques for non-destructive and destructive examination of nuclear fuel systems (e.g. covering fuel and cladding or fuel and coatings, or other configurations), 'in-depth' methods are applied to determine physical, thermomechanical and micro-analytical properties of specific phases/features present in irradiated fuel. This paper presents relevant trends and examples from fuel safety studies performed at JRC-Karlsruhe and in other laboratories which highlight the 'added value' obtainable by applying combined techniques, high resolution tools or 3D methods of analysis for the experimental 'multiscale' characterization of irradiated fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Assessing High Burnup U-19Pu-10Zr Fuel Performance against Historical and Modeled Behavior

Advancing the deployment of sodium-cooled fast reactors (SFRs) requires thorough testing of metallic fuel pins under accident conditions to establish safe operational limits of high burnup fuel. To conduct transient testing, a comprehensive understanding of steady-state fuel behavior obtained through both experimental characterization and accurate predictive capabilities is needed. This study comparatively assesses the steady-state irradiation performance of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, irradiated under prototypic fast reactor conditions in preparation for planned safety testing at the Transient Reactor Test Facility. Since DP-40 was designated for use in the test and DP-36 serves as its sibling pin, non-destructive, engineering-scale post-irradiation examinations (PIE) were conducted on both pins while destructive examinations were performed exclusively on DP-36. The results were then assessed against historical performance data from similar fuel pins irradiated in the Experimental Breeder Reactor-II. Additionally, the steady-state irradiation of each pin was modeled using the BISON fuel performance code to assess the accuracy of current modeling capabilities in predicting the baseline irradiation behavior. Non-destructive examinations included neutron radiography to measure fuel column elongation, gamma scanning to verify pin integrity and fission product migration, and profilometry to assess dimensional changes. Benchmarking against existing PIE data revealed consistent patterns in axial fuel column growth and cladding diametral strain, though both pins exhibited longer low-density “fluff” structures, which can have implications for core reactivity and source term calculations. Destructive examinations on DP-36 included fission gas release analysis and sectioning for optical microscopy, which showed more complex constituent redistribution patterns than the traditionally accepted 3-ring model. The axial evolution of fractional areas and porosities of each of the redistributed zones were quantified and presented. Modeling comparisons showed agreement in fractional fission gas release but consistently overestimated axial and radial swelling and disagreed with measured axial porosity patterns. These conservative overpredictions suggested that the pins would appear closer to failure or operational limits at the start of transient tests, potentially leading to higher strain accumulation during the transient. While conservative estimates provide safety margins, they can negatively impact fuel economics. A review of the swelling models identified areas for improvement in the gaseous swelling, solid swelling, and fuel hot-pressing models when applied to ternary fuel. The results of this study highlight the critical importance of conducting pre-test characterization on both test and sibling pins to accurately capture steady-state fuel behavior, providing a precise baseline for post-test evaluations and essential inputs for transient modeling of the planned experiments. The analysis also revealed significant data gaps that require further investigation to enhance the understanding and prediction of fuel swelling and pore dynamics. Collecting comprehensive data across different irradiation conditions, burnup levels, and fuel compositions are essential for refining existing models and developing mechanistic models for both binary and ternary metallic fuels, ultimately improving the integration of modeling and experimental approaches in accident testing.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗