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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 415 records · Page 23
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Sequences in Linker-1 domain of the multidrug resistance associated protein (MRP1 or ABCC1) bind to tubulin and their binding is modulated by phosphorylation

The multidrug resistant associated protein 1 (or MRP1, ABCC1) encodes two cytoplasmic linker domains (L0 and L1) composed of highly charged sequences with multiple protein kinase A/C phosphorylation sites. In this report we made use of the scanning peptide approach to identify MRP1 linker L1 (L1{sup MRP1}) interacting proteins. Scanning heptapeptides covering L1{sup MRP1} 126 amino acids (Ile{sup 846}- Leu{sup 972}) were synthesized and used in pull-down assays to isolate proteins from cell extracts (human multidrug resistant SCLC cell line; H69/AR). The results show four high affinity binding sequences in L1{sup MRP1} domain [{sup 866}FLRTYAST{sup 867}; {sup 906}SAGKQLQRQLSSS{sup 912}; {sup 925}ISRHHNSTA{sup 927} and {sup 954}AQTGQVKLSVYW{sup 959}] that bound ∼55 kDa and 110 kDa polypeptides. The latter polypeptides were identified by mass spectrometry as α- and β-tubulin monomers and dimers. Western blotting with monoclonal antibodies to α- and β-tubulin proteins confirmed the mass-spectrometry results. Moreover, using recombinant full-length GST-Linker 1 fusion polypeptide (GST-L1{sup MRP1}), we confirmed the peptide scanning approach demonstrating specific binding of tubulin to GST-L1{sup MRP1}. Intriguingly, substitutions of serine residues in L1{sup MRP1} by aspartic acid, but not alanine, abolished its binding to tubulin, suggesting that phosphorylation of Ser{sup 871,} {sup 915,} {sup 930,} {sup and} {sup 961} within L1{sup MRP1} may modulate its binding to tubulin. Taken together, the results of this study suggest possible interaction between MRP1 and tubulin that is modulated by phosphorylation of specific sequences in the L1{sup MRP1} domain. - Highlights: • This study shows direct binding between MRP1 Linker 1 (L1{sup MRP1}) domain and α- and β-tubulin subunits, using a scanning peptide approach. • Four short and dispersed sequences in L1{sup MRP1} domain interact with α/β-tubulin subunits. • Each sequence in L1{sup MRP1}-tubulin binding domain contains predicted protein kinase C/A site. • Substituting 3 of 4 serine with aspartic acid, but not alanine, in L1{sup MRP1} binding domains inhibit their binding to tubulin subunits.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2017

Engineered antibody CH2 domains binding to nucleolin: Isolation, characterization and improvement of aggregation

Smaller recombinant antibody fragments are now emerging as alternatives of conventional antibodies. Especially, immunoglobulin (Ig) constant CH2 domain and engineered CH2 with improved stability are promising as scaffolds for selection of specific binders to various antigens. We constructed a yeast display library based on an engineered human IgG1 CH2 scaffold with diversified loop regions. A group of CH2 binders were isolated from this yeast display library by panning against nucleolin, which is a tumor-associated antigen involved in cell proliferation, tumor cell growth and angiogenesis. Out of 20 mutants, we selected 3 clones exhibiting relatively high affinities to nucleolin on yeasts. However, recombinant CH2 mutants aggregated when they were expressed. To find the mechanism of the aggregation, we employed computational prediction approaches through structural homology models of CH2 binders. The analysis of potential aggregation prone regions (APRs) and solvent accessible surface areas (ASAs) indicated two hydrophobic residues, Val{sub 264} and Leu{sub 309}, in the β-sheet, in which replacement of both charged residues led to significant decrease of the protein aggregation. The newly identified CH2 binders could be improved to use as candidate therapeutics or research reagents in the future. - Highlights: • Novel engineered CH2 domains to nucleolin were identified from the yeast display CH2-based library. • Aggregation prone regions (APRs) were predicted through structural homology models CH2 domains. • The replacement of two hydrophobic residues in APRs significantly led to decrease of the CH2 aggregation.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2018

Structural and biochemical studies on the role of active site Thr166 and Asp236 in the catalytic function of D-Serine deaminase from Salmonella typhimurium

Highlights: • The crystal structure of T166A determined at 2.7 Å resolution represents the first PLP bound form of the StDSD enzyme. • Mutation of Thr166 to Ala and Asp236 to Leu results in the inactivation of the enzyme. • Thr166 might abstract Cα proton of D-Ser to carry out catalysis and Asp236 might stabilize the internal aldimine form. • Conformational changes are important for bringing Thr166 to a position favorable for proton abstraction. D-Serine deaminase (DSD) degrades D-Ser to pyruvate and ammonia. Uropathogenic bacteria survive in the toxic D-Ser containing mammalian urine because of DSD activity. The crystal structure of the apo form of Salmonella typhimurium DSD (StDSD) has been reported earlier. In the present work, we have investigated the role of two active site residues, Thr166 and Asp236 by site directed mutagenesis (T166A and D236L). The enzyme activity is lost upon mutation of these residues. The 2.7 Å resolution crystal structure of T166A DSD with bound PLP reported here represents the first structure of the holo form of StDSD. PLP binding induces small changes in the relative dispositions of the minor and major domains of the protein and this inter-domain movement becomes substantial upon interaction with the substrate. The conformational changes bring Thr166 to a position at the active site favorable for the degradation of D-Ser. Examination of the different forms of the enzyme and comparison with structures of homologous enzymes suggests that Thr166 is the most probable base abstracting proton from the Cα atom of the substrate and Asp236 is crucial for binding of the cofactor.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2018

A composite position independent monitor of reactor fuel irradiation using Pu, Cs, and Ba isotope ratios

When post-irradiation materials from the nuclear fuel cycle are released to the environment, certain isotopes of actinides and fission products carry signatures of irradiation history that can potentially aid a nuclear forensic investigation into the material's provenance. In this study, combinations of Pu, Cs, and Ba isotope ratios that produce position (in the reactor core) independent monitors of irradiation history in spent light water reactor fuel are identified and explored. These position independent monitors (PIMs) are modeled for various irradiation scenarios using automated depletion codes as well as ordinary differential equation solutions to approximate nuclear physics models. Experimental validation was performed using irradiated low enriched uranium oxide fuel from a light water reactor, which was sampled at 8 axial positions from a single rod. Plutonium, barium and cesium were chemically separated and isotope ratio measurements of the separated solutions were made by quadrupole and multi-collector inductively coupled mass spectrometry (Cs and Pu, respectively) and thermal ionization mass spectrometry (Ba). The effect of axial variations in neutron fluence and energy spectrum are evident in the measured isotope ratios. Furthermore, two versions of a combined Pu and Cs based PIM are developed. A linear PIM model, which can be used to solve for irradiation time is found to work well for natural U fuel with <10% 240 Pu and known or short cooling times. A non-linear PIM model, which cannot be solved explicitly for irradiation time without additional information, can nonetheless still group samples by irradiation history, including high burnup LEU fuel with unknown cooling time. 137 Ba/ 138 Ba is also observed to act as a position independent monitor; it is nearly single valued across the sampled fuel rod, indicating that samples sharing an irradiation history (same irradiation time and cooling time) in a reactor despite experiencing different neutron fluxes will have a common 137 Ba/ 138 Ba ratio. Modeling of this Ba PIM shows it increases monotonically with irradiation and cooling time, and a confirmatory first order analytical solution is also presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Pore growth in U-Mo/Al dispersion fuel

U-Mo/Al dispersion fuel is currently under development in the DOE’s Material Management and Minimization program to convert HEU-fueled research reactors to LEU-fueled reactors. In some demanding conditions in high-power and high-performance reactors, large pores form in the interaction layers between the U-Mo fuel particles and the Al matrix, which pose a potential to cause fuel failure. In this study, comprehension of the formation and growth of these pores was explored. As a product, a model to predict pore growth and porosity increase was developed. Well-characterized in-pile data from reduced-size plates were used to fit the model parameters. A data set of full-sized plates, independent and distinctively different from those used to fit the model parameters, was used to examine the accuracy of the model.

in-pile data↗
DOE OSTI2017

Microstructural characterization of an irradiated RERTR-6 U-7Mo/AA4043 alloy dispersion fuel plate specimen blister-tested to a final temperature of 500°C

The Material Management and Minimization (M3) Reactor Conversion Program, in the past called the Reduced Enrichment for Research and Test Reactor (RERTR) Program, is developing low-enriched uranium (LEU) fuels for application in research reactors. U–Mo alloy dispersion fuel is one type being developed. Blister testing has been performed on different fuel plate samples to determine the margin to failure for fuel plates irradiated to different fission densities. Microstructural characterization was performed using scanning electron microscopy and transmission electron microscopy on a sample taken from a U-7Mo/AA4043 matrix dispersion fuel plate irradiated in the RERTR-6 experiment that was blister-tested up to a final temperature of 500°C. The results indicated that two types of grain/cell boundaries were observed in the U- 7Mo fuel particles, one with a relatively low Mo content and fission gas bubbles and a second type enriched in Si, due to interdiffusion from the Si-containing matrix, with little evidence of fission gas bubbles. With respect to the behavior of the major fission gas Xe, a significant amount of the Xe was still observed within the U-7Mo fuel particle, along with microns into the AA4043 matrix. For the fuel/matrix interaction layers that form during fabrication and then grow during irradiation, they change from the as-irradiated amorphous structure to one that is crystalline after blister testing. In the AA4043 matrix, the original Si-rich precipitates, which are typically observed in as-irradiated U-Mo dispersion fuel, get consumed due to interdiffusion with the U-7Mo fuel particles during the blister test. Lastly, the fission gas bubbles that were originally around 2 nm in diameter and resided on a fission gas superlattice in the intragranular regions of as-irradiated U-7Mo fuel grew in size (up to ~20 nm diameter) during blister testing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Non-destructive analysis of swelling in the EMPIrE fuel test

The European Mini-Plate Irradiation Experiment (EMPIrE) was designed to support the development and testing of a coated uranium-molybdenum (U-Mo) dispersion fuel for the conversion of select high-performance research reactors (HPRRs) to utilize low-enriched uranium (LEU). To aid in the development of the coated fuel form, the EMPIrE test included several plate designs and irradiated them in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) at a high meat power density (~21 kW/cm 3 ) and to high fuel particle fission densities (~6.4 × 10 21 fissions/cm 3 ). These conditions mimic the bounding conditions of the BR-2 reactor in Belgium, where a concurrent irradiation experiment was performed, and exceed those previously explored in dispersion U-Mo fuel plates. A local fuel swelling analysis, as determined through high-fidelity, post-irradiation mini-plate profilometry, was used along with statistical methods to non-destructively evaluate the overall performance and separate the effects of convoluted fabrication variables. While some effects observed with this non-destructive analysis were subtle, others had more significant, and possibly competing, effects on the fuel swelling behavior. In closing, these observations will be examined further with destructive examinations to more fully assess them as the fuel design is developed and qualified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A new method to measure the U-235 content in fresh LWR fuel assemblies via fast-neutron passive self-interrogation

The purpose of this paper is to provide a solution to a decades old safeguards problem in the verification of the fissile concentration in fresh light water reactor (LWR) fuel assemblies. The problem is that the burnable poison (e.g. Gd 2 O 3 ) addition to the fuel rods decreases the active neutron assay for the fuel assemblies. This paper presents a new innovative method for the verification of the 235 U linear mass density in fresh LEU fuel assemblies that is insensitive to the burnable poison content. The technique makes use of the 238 U atoms in the fuel rods to self-interrogate the 235 U mass. The innovation for the new approach is that the 238 U spontaneous fission (SF) neutrons from the rods induces fission reactions (IF) in the 235 U that are time correlated with the SF source neutrons. Thus, the coincidence gate counting rate benefits from both the nu-bar of the 238 U SF (2.07) and the 235 U IF (2.44) for a fraction of the IF reactions. Whereas, the 238 U SF background has no time-correlation boost. The higher the detection efficiency, the higher the correlated boost because background neutron counts from the SF are being converted to signal doubles. This time-correlation in the IF signal increases signal/background ratio that provides a good precision for the net signal from the 235 U mass. The hard neutron energy spectrum makes the technique insensitive to the burnable poison loading where a Cd or Gd liner on the detector walls is used to prevent thermal-neutron reflection back into the fuel assembly from the detector. Here, we have named the system the fast-neutron passive collar (FNPC).

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2026

The porosity surrounding carbides and second phase stringers in monolithic U-10Mo fuel plate after irradiation

Post-irradiation microstructure characterization plays an important role in qualifying the low-enriched uranium (LEU) monolithic U-10 wt%Mo plate-type fuel for United States high-performance research reactors (USHPRRs) program. Inhomogeneous features resulting from manufacturing and irradiation processes, including carbides, second phase stringers, and extensive void spaces caused by the combining of small porosities, may increase the risk of heat concentration in local regions of the fuel plate over the operating conditions. In this study, characteristics of carbides, stringers, and porosity after multiple levels of irradiation at varying fission densities were studied by electron microscopes to decipher the morphology of pores and the porosity evolution in U-10 wt%Mo. For carbides, the result shows that porosities start forming on UMo grain boundaries, then on UMo/carbides interfaces as the burn-up going higher. However, the porosities surrounding carbides grow larger than the ones on UMo grain boundaries. The porosities around the uranium carbides could interconnect to form larger void space. The study revealed that the void spaces larger than 5 µm were found around uranium carbides after high burnup, while no evidence was observed to support the similar voids formed near second phase stringers even though the size of the stringers (> 50 µm) was much larger than uranium carbides (< 20 µm). The evolution of porosities suggests that the formation of second phase stringers may not create more significant porosities compared to regular uranium carbides regions during fuel operating conditions.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Status of the NGNP fuel experiment AGR-2 irradiated in the advanced test reactor

The United States Department of Energy's Next Generation Nuclear Plant (NGNP) Advanced Gas Reactor (AGR) Fuel Development and Qualification Program will be irradiating up to seven separate low enriched uranium (LEU) tri-isotopic (TRISO) particle fuel (in compact form) experiments in the Advanced Test Reactor (ATR) located at the Idaho National Laboratory (INL). These irradiations and fuel development are being accomplished to support development of the next generation reactors in the United States, and will be irradiated over the next several years to demonstrate and qualify new TRISO coated particle fuel for use in high temperature gas reactors. The goals of the irradiation experiments are to provide irradiation performance data to support fuel process development, to qualify fuel for normal operating conditions, to support development and validation of fuel performance and fission product transport models and codes, and to provide irradiated fuel and materials for post irradiation examination (PIE) and safety testing. The experiments, which will each consist of at least six separate capsules, will be irradiated in an inert sweep gas atmosphere with individual on-line temperature monitoring and control of each capsule. The sweep gas will also undergo on-line fission product monitoring to track performance of the fuel in each individual capsule during irradiation. The first experiment (designated AGR-1) started irradiation in December 2006 and was completed in November 2009. The second experiment (AGR-2), which utilized the same experiment design as well as control and monitoring systems as AGR-1, started irradiation in June 2010 and is currently scheduled to be completed in April 2013. The design of this experiment and sup

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Development of a thermal creep model for aluminum alloy 6061 cladding in U-10Mo monolithic fuel plates

Plate-type fuel elements consisting of a high-density, low-enriched uranium (LEU) U–10Mo-based fuel foil encapsulated in an aluminum alloy (AA) cladding are fabricated using the hot isostatic pressing (HIP) technique. During the HIP process, the fuel plate system is heated to 560 °C, then cooled to room temperature. This heat cycle significantly affects the mechanical properties of the aluminum cladding, and experimental investigations have shown that, post-HIP bonding, the mechanical properties of the aluminum cladding transition from those of AA 6061-T6 to something closer to the O temper. More specifically, the ultimate strength of the cladding decreases while its ductility increases, making it challenging to capture the changes in mechanical behavior and material properties. Understanding the residual stresses generated during the HIP process is critical for assessing the fuel plate’s integrity under various temperature, pressure, and irradiation. To simulate the HIP bonding process, the elastic, plastic, and thermal properties of the cladding are assumed to be similar to those of AA 6061-O temper. However, the primary challenge lies in the lack of available data for the creep model of the AA 6061 cladding during this transient process of HIP. The present study focuses on developing a computational model that predicts the creep behavior of the aluminum cladding in the fuel plates during the HIP process, as cladding creep significantly influences the residual stresses generated in U-10Mo fuel plates during HIP fabrication. Furthermore, as HIP bonding occurs at high temperatures that are nearing the melting point of aluminum, the present work considered a temperature-dependent Arrhenius-type creep model. In particular, a hyperbolic sine creep model is employed to estimate the creep properties of the as-fabricated aluminum cladding. In conclusion, the residual stresses predicted in the U-10Mo fuel when using the newly calibrated creep model closely align with the experimental measurements, validating the model’s accuracy.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Forced flow transient safety analysis of irradiation device with adjustable orifice for research reactor fuel assemblies

The Belgium Reactor 2 (BR2) of the Belgian Nuclear Research Centre (SCK CEN) has several irradiation devices or rigs that are dedicated to the fuel performance and qualification demonstration testing of research reactor fuels. In support of the U.S. High Performance Research Reactor (USHPRR) LEU conversion project, a new flexible irradiation apparatus, MUSTANG-R, has been constructed. SCK CEN has completed the design and safety study, in cooperation with Idaho National Laboratory (INL) and Argonne National Laboratory (ANL), to allow for the irradiation testing of a full-size fuel assembly in a 200 mm diameter channel in the BR2 reactor. The moveable valve is a key design feature of the device and acts like an adjustable orifice enhancing or restricting the flow through a coolant channel inlet located in the BR2 upper plenum. This moveable valve allows the flow through the device to be adjusted prior to each BR2 cycle to obtain the necessary conditions for the fuel qualification test. This ensures accurate and representative thermal-hydraulic conditions of the fuel design are achieved. The device was designed and qualified as passively safe, implying verification by a combination of mechanical and thermal-hydraulic analysis and testing. This includes characterization of the safety margin required for a scenario where the moveable valve is assumed to be erroneously closed during irradiation. A simplified and conservative method is proposed for analyzing the corresponding forced flow transient using a critical heat flux criterion. In conclusion, this allows the required minimum valve opening to be determined for the experiments' design and safety studies.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Development and biodistribution studies of 77 As-labeled trithiol RM2 bioconjugates for prostate cancer: Comparison of [ 77 As]As-trithiol-Ser-Ser-RM2 vs. [ 77 As]As-trithiol-Glu-Ser-RM2

Recent progress with the production of 72 As (2.49 Mev β + max (64%), 3.33 Mev β + max (16%), 834 keV (81%), t 1/2 : 26 h) and 77 As (0.683 Mev β- max (97%), 239 keV (1.59%), t 1/2 : 38.8 h) has facilitated their evaluation as a potential “theranostic pair” for PET imaging and radiotherapy. Our 3 rd generation trithiol chelate with two carboxylic acid groups was further developed as a bifunctional chelate for radioarsenic. The As complex with the trithiol chelate was synthesized and characterized. No carrier added (nca) [ 77 As][H 2 AsO 4 - ] was used for radiolabeling studies. Here, the trithiol chelate was conjugated to the RM2 peptide (DPhe-Gln-Trp-Ala-VaI-Gly-His-Sta-Leu-NH 2 ) via solid phase peptide synthesis with two different linkers, Ser-Ser and Glu-Ser. The trithiol chelate and its RM2 bioconjugates were radiolabeled with nca 77 As, and the RM2 bioconjugates were compared in initial biodistribution studies. The As diacid trithiol complex was characterized by 1 H NMR, 13 C NMR and HR-ESI-MS. The trithiol-RM2 precursor and As trithiol bioconjugates were characterized by HR-ESI-MS and/or LC-ESI-MS. Radiolabeling of the RM2 bioconjugates with 77 As resulted in over 85% radiochemical yield for [ 77 As]As-trithiol-Ser-Ser-RM2 ([ 77 As]8) and 90% for [ 77 As]As-trithiol-Glu-Ser-RM2 ([ 77 As]9). Both radiotracers demonstrated excellent in vitro stability (≥ 90% remaining intact through 24 h in PBS buffer) and were more hydrophilic than previous analogues based on log D 7.4 values. Biodistribution results of the two radiotracers in healthy CF-1 male mice demonstrated blockable pancreatic uptake at 1 h (82% for ([ 77 As]8 and 78% for [ 77 As]9) indicating specific gastrin-releasing peptide receptor (GRPR) uptake. The primary route of excretion was through the gastrointestinal system for both radiotracers. A new trithiol chelate with improved hydrophilicity was successfully conjugated to the RM2 peptide via two linkers, and high radiolabeling yield with nca 77 As was achieved. In vivo biodistribution studies with both radiotracers demonstrated blockable pancreatic uptake suggestive of specific receptor uptake.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

Impact of recent ENDF nuclear data update, high initial enrichment and high burnup fuel on critical experiments applicability determination via the integral index c k for burnup credit validation

In 2012, NUREG/CR-7109 reported on the validation of burnup credit calculations involving major and minor actinides and major fission products which was investigated for pressurized and boiling water reactor (PWR and BWR) fuel enrichments up to 5 wt% 235 U and assembly-average burnups up to 60 GWd/MTU. Recently, there has been interest in increasing the maximum enrichment used in PWR fuel as high as 8 wt% 235 U and correspondingly increasing the maximum assembly-average burnups to approximately 75 GWd/MTU. These proposed increases in enrichment and burnup necessitate reinvestigation of the validation basis for k eff calculations for this expanded application space. Additionally, the 2012 study was performed by using the Evaluated Nuclear Data File (ENDF)/B-VII.0 nuclear data with the SCALE 6 covariance library, and the effects of using the newly released ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data and covariance libraries should be evaluated. In this work, published in NUREG/CR-7309 in 2025, the validation assessment was performed consistently with NUREG/CR-7109: modeling irradiated fuel assemblies in the Generic Burnup Credit (GBC)-32 cask defined in NUREG/CR-6747. The TSUNAMI-3D sequence was used to generate sensitivity data for the application model, and the data were compared with sensitivity data from select benchmark models. The integral parameter c k is the metric of similarity used in this study and is consistent with NUREG/CR-7109, where a c k value in excess of 0.8 indicates sufficient similarity for use in validation. A new set of benchmark experiments with sensitivity data has been assembled for this effort. The number of experiments with available sensitivity data is now 2,104, compared to 474 in NUREG/CR-7109. This increase was facilitated by the efforts of the Nuclear Energy Agency to generate sensitivity data for a majority of the experiments in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook to supplement the data available in the Oak Ridge National Laboratory (ORNL) Verified, Archived Library of Inputs and Data (VALID). The complete set of benchmarks considered here includes experiments for low-enriched uranium (LEU), intermediate enriched uranium (IEU), and a mixture of uranium and plutonium (MIX) from the ICSBEP Handbook and VALID, as well as ORNL models of the Haut Taux de Combustion (HTC) experiments and other potentially relevant models not included in VALID. The updated similarity study shows that none of the extended burnup and higher enrichment combinations considered show a significant decrease in the number of potentially applicable experiments, meaning sufficient critical experiments exist for the validation of BUC criticality safety calculations, with initial enrichments up to 8 wt% 235 U and burnups up to 80 GWd/MTU. Additionally, both the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries can be used for validation since the number of critical experiments applicable for validation increases for most cases with the most recent nuclear data compared to the previous one. As in previous BUC validation studies, the French HTC experiments are the most similar in a majority of the application cases studied, especially from representative discharge burnups ranging from 40 to 80 GWd/MTU. In conclusion, these results match the conclusions presented in NUREG/CR-7109 regarding validation of the primary actinides in BUC analyses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Nano-Scale Fission Product Phases in an Irradiated U-7Mo Alloy Nuclear Fuel

Irradiated nuclear fuel is a very difficult material to characterize. Due to the large radiation fields associated with these materials, they are hard to handle and typically have to be contained in large hot cells. Even the equipment used for performing characterization is housed in hot cells or shielded glove boxes. The result is not only a limitation in the techniques that can be employed for characterization, but also a limitation in the size of features that can be resolved The most standard characterization techniques include light optical metallography (WM), scanning electron microscopy (SEM), and electron probe microanalysis (EPMA). These techniques are applied to samples that are typically prepared using grinding and polishing approaches that will always generate some mechanical damage on the sample surface. As a result, when performing SEM analysis, for example, the analysis is limited by the quality of the sample surface that can be prepared. However, a new approach for characterizing irradiated nuclear fuel has recently been developed at the Idaho National Laboratory (INL) in Idaho Falls, Idaho. It allows for a dramatic improvement in the quality of characterization that can be performed when using an instrument like an SEM. This new approach uses a dual-beam scanning microscope, where one of the beams isa focused ion beam (FIB), which can be used to generate specimens of irradiated fuel (-10µm x 10µm) for microstructural characterization, and the other beam is the electron beam of an SEM. One significant benefit of this approach is that the specimen surface being characterized has received much less damage (and smearing) than is caused by the more traditional approaches, which enables the imaging of nanometer­ sized microstructural features in the SEM. The process details are for an irradiated low-enriched uranium (LEU) U-Mo alloy fuel Another type of irradiated fuel that has been characterized using this technique is a mixed oxide fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Rapid Droplet Sampling Interface for Low-Volume, High-Throughput Mass Spectrometry Analysis

Here, we present a rapid droplet sampling interface (RDSI) electrospray ionization mass spectrometry (ESI-MS) system as a high-throughput, low-volume, noncontact, and minimal-carryover approach for characterization of liquids. Liquid characterization was achieved by combining droplet ejection with an open-face microflow capillary with a 2.5 μL/min continuous flow of carrier solvent. Through this implementation, single 0.3 nL droplets containing the analyte effectively mix with 4–8 nL of carrier solvent and create a combined electrospray plume. The carrier solvent continuously cleaned the system, eliminating carryover. A sampling rate of 5 Hz was achieved for droplets containing 1 μM propranolol or 5 μM leu-enkephalin with each droplet fully baseline-resolved (138 ± 32 ms baseline-to-baseline). Using a SCIEX API4000 mass spectrometer, a lower limit of quantification (LLOQ) of propranolol was 15 nM, corresponding to 1.16 fg of propranolol in the droplet, and was linear across 3 orders of magnitude. Further, quantitation could be achieved by adding an isotopically labeled internal standard, as done in conventional ESI. Signal transients were faster than the acquisition speed of the mass spectrometer, resulting in artificially high reproducibility of 15–30% RSD droplet-to-droplet. Analyte–solvent mixing ratios could be controlled by adjusting droplet positioning along the open-face capillary with an optimal position about 0.4 mm from the tip end. The range of analyte coverage was exemplified by measures of peptides and drugs in methanol, water, and buffer solutions. In a comparison to the Open Port Sampling Interface (OPSI) implemented on the same system, the RDSI had 78× greater sensitivity, 6× greater throughput and used significantly less carrier solvent.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗