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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 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Nuclide inventory validation: radiochemical assay data quality and modeling challenges in benchmark models development

Oak Ridge National Laboratory is conducting radiochemical assay experiments, using high-precision analytical protocols validated with a comprehensive quality assurance plan, to expand the nuclide inventory validation basis for high burnup spent nuclear fuel. Preliminary measurement data for key actinides and fission products in two pressurized water reactor spent fuel samples are being used to investigate the impact of measurement data uncertainty on the sample burnup estimation. These measurement data are also being used to examine the impact of assumptions applied when developing best-estimate computational models to simulate fuel irradiation history. The simulations are being performed using depletion capabilities in the SCALE code system. Comparison of calculated and measured nuclide concentrations shows good agreement for the considered nuclides. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Behavior of the Mo, Tl, and U isotope systems during differentiation in the Kilauea Iki lava lake

Stable molybdenum (Mo), thallium (Tl), and uranium (U) isotope ratios were determined in a suite of samples from the 1959 Kilauea eruption and from Kilauea Iki lava lake with the aim of understanding the effects of igneous differentiation on these isotope systems. The samples range from olivine cumulate with MgO up to 27% to internal differentiates with MgO less than 3%, representing a tholeiitic differentiation series. Molybdenum, Tl, and U behave incompatibly during differentiation, and Mo and U isotope ratios do not systematically vary amongst the different samples. δ 98 Mo values range from -0.17 to -0.31‰ and δ 238 U values range from -0.20 to -0.38‰. Most individual analyses for both isotope systems overlap within measurement uncertainty (± ~0.7 and ~ 0.6, respectively). Mean δ 98 Mo and δ238U values are -0.22 ± 0.08‰ (2σ) and - 0.29 ± 0.09‰ (2σ), respectively, which overlap with Pacific mid ocean ridge basalt (MORB). In contrast, Tl isotopes show small but resolvable variations, with ε 205 Tl ranging from +1.20 to -1.38. The most negative ε 205 Tl values are confined to some of the lowest [Tl] samples, but the ε 205 Tl values do not otherwise vary smoothly with MgO or [Tl]. Possible mechanisms for thallium isotope fractionation are considered (e.g., degassing, water leaching, sulfide fractionation) but none are found to be satisfactory. Overall, the lack of resolvable variation in the Mo and U isotope systems and the small magnitude of heterogeneity in the Tl isotope system indicate that differentiation in tholeiitic systems is unlikely to be a major contributor to global variation in these isotope systems.

58 GEOSCIENCES↗
DOE OSTI2023

Protecting People and Planet

To produce energy, civilian nuclear power plants require fuel. The majority of existing plants rely on solid ceramic fuels that contain concentrations of about 3 to 5% of the element uranium-235. This specific isotope is well suited to sustaining the type of chain reaction required for a nuclear reactor to harness the heat produced by nuclear fission. Title I of the United States Atomic Energy Act of 1954 defines “special nuclear material” as plutonium, uranium-233, or uranium enriched in the isotopes uranium-233 or uranium-235. This label identifies materials that at high concentrations can be used to produce nuclear weapons. The United States and broader international community are fully committed to ensuring the peaceful use of nuclear energy by committing to practices known as safeguards.

99 GENERAL AND MISCELLANEOUS↗
NASA NTRS1977

Lead isotope studies of mare basalt 70017

Uranium, thorium, and isotopic lead data for components of basalt 70017 are reported, and it is found that the whole rock, pyroxene, and ilmenite points in a concordia diagram plot along a chord intersecting the curve at 3.7 and 4.33 eons. The plagioclase data do not seem to lie on this line. The data for 70017 appear to plot along a distinctly different chord in a concordia diagram than do the data for 75055 and 75035, two other Apollo 17 mare basalts. The lead data are in accord with Sm-Nd results. A 3.7 eon crystallization age for 70017 would be consistent with the same kind of parentless lead that is indicated by previous studies of soils and soil breccias from stations at Taurus-Littrow. The Th/U ratio in ilmenite is 2.2, and the concentrations of these two elements are approximately twice those in pyroxene.

Mattinson, J. M.↗
NASA NTRS1978

Lead isotope systematics of mare basalt 75075

Uranium, thorium and isotopic lead data are reported for two bulk samples and separated pyroxene, ilmenite and plagioclase from basalt 75075. In a concordia diagram the whole rock, ilmenite and four pyroxene samples define a chord intersecting the concordia curve at approximately 4.25 and 2.8 AE. Three plagioclase samples plot distinctly off the chord. The crystallization age of 75075 is accurately determined at 3.74 AE by Rb-Sr, Sm-Nd and K-Ar measurements from other laboratories. It is not possible to adjust the isotopic composition of initial lead so as to reconcile the U-Pb data with a crystallization age of 3.74 AE. The data therefore indicate some type of post-crystallization disturbance of the U-Pb system that is not detected by the other systems. The 75075 data are one of the few examples of this type of age pattern found on the moon. If the disturbance was a single event, it probably occurred around 2.8 AE ago, the time indicated by the pyroxene, whole rock and ilmenite data.

Chen, J. H.↗
DOE OSTI2025

High Flux Isotope Reactor Low-Enriched Uranium High Density Silicide Fuel Preliminary Design Update: System Transient Analysis

As a part of conversion efforts from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel under direction of the National Nuclear Security Administration of the U.S. Department of Energy, multiple proposed designs of the High Flux Isotope Reactor (HFIR) have been created and assessed regarding reactor physics performance metrics, including designs utilizing uranium silicide dispersion fuel (U3Si2-Al). This report updates the previous analyses that evaluated the nuclear safety performance of LEU fuel designs with respect to selected accident events from the HFIR Safety Analysis Report (SAR). Both the Low Density (LD) and High Density (HD) Optimized designs’ reactivity initiated accident fuel performance improved relative to the HEU fuel, attributed to greater 238 U negative Doppler feedback. However, the thermal margins for primary coolant system accidents were reduced with some acceptance criteria unable to be met. The need to resolve reduced thermal margin, open modeling items, and unresolved assumptions was identified.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Producing 236 U reference standards for Accelerator Mass Spectrometry at the University of Notre Dame

36 U is a rare isotope of uranium, naturally occurring in ores with an abundance of 236 U/ 238 U$<$ 1 x 10 -9 . The ability to detect it and make isotopic ratio measurements has applications ranging from nuclear forensics and nonproliferation to energy production and environmental protection. Currently, Accelerator Mass Spectrometry (AMS) is the only technique sensitive enough to accurately measure 236 U/ 238 U isotopic ratios as they exist in naturally occurring ores in the range of 236 U/ 238 U = 10 -12 $-$ 10 -9 . Some AMS facilities have demonstrated their capabilities to make these measurements. Historically, the lack of commercially available reference standards covering the range of naturally occurring 236 U/ 238 U abundances has necessitated the use of absolute measurements, notoriously difficult to do using AMS, resulting in increased uncertainties in measurements and a reliance on knowledge of systematic effects. To mitigate these issues, various AMS facilities have sought to develop their own reference standards. Using a reference standard prepared for other forms of mass spectrometry, a series of AMS suitable standards was created through dilution with low-background natural uranium. The techniques used to produce and characterize these materials as well as analysis of them using AMS will be discussed.

236U↗
DOE OSTI2021

Exploring the use of thorium isotope compositions and concentrations as nuclear forensic signatures for uranium ore concentrates

This interlaboratory study measured thorium concentrations and isotope compositions in uranium ore concentrates from different geographical locations to examine whether thorium impurities may be useful forensic signatures for uranium ore concentrates found out of regulatory control. Measured 230 Th/ 232 Th in fifteen uranium ore concentrates record over three orders of magnitude of compositional variation. Results demonstrate that 230 Th/ 232 Th used in combination with U/Th ratios resulted in a unique signature for individual uranium ore concentrates from different processing locations. Data presented here suggest potential for 230 Th/ 232 Th and U/Th to be used as comparative signatures to investigate the provenance of seized uranium ore concentrates.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2025

Precision Local Burnup Assessment Through Dynamic Peak Fitting in Atom Probe Tomography for Depleted, Enriched, and Irradiated Metallic and Ceramic Fuels

Abstract Burnup estimation in nuclear fuels is vital for evaluating fuel performance, transportation, and safe fuel storage. Accurate assessments of burnup from service period and spent fuels involve tracking the consumption of fissile isotopes of uranium (U) offering a direct insight into energy changes within the fuels especially for thermal spectrum reactors. In current approach, mass spectroscopic technique in atom probe tomography (APT) is utilized for accurate quantification of U isotopes. Quantification of U peaks in mass spectrum is performed on asymmetric shapes due to delayed signals, known as thermal tails, particularly for poorly conducting samples analyzed in laser mode. In this study, we introduce a novel quantification tool for isotopic analysis from APT datasets by developing a fitting algorithm based on shapes of the peaks. A MATLAB-based dynamic peak fitting toolbox is developed and designed to adapt to various peak shapes, ensuring accurate quantification of U isotopes. The effectiveness of this approach is demonstrated in standard Ni-Cr sample, depleted and enriched U samples, and U-based fuels with different burnup levels. The viability of this approach for isotopic quantification is demonstrated on both metallic and ceramic fuels.

Burnup↗
DOE OSTI2020

Conceptual Fuel Element Design Candidates for Conversion of High Flux Isotope Reactor with Low-Enriched Uranium Silicide Dispersion Fuel

Engineering design studies are underway to assess the feasibility of converting the High Flux Isotope Reactor (HFIR) to operate with low-enriched uranium (LEU) silicide dispersion (U3Si2-Al) fuel. These studies are supported by the U.S. Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization. A systematic approach employing neutronic and thermal-hydraulic analyses have been performed with the ORNL Shift and HFIR Steady State Heat Transfer Code tools, respectively, to predict reactor performance and thermal safety margins for proposed LEU3Si2-Al fuel designs. The design process was initiated by generating an optimized design with fabrication features identified from previous studies that result in excellent performance and safety metrics. The approach continued by substituting a single fabrication feature anticipated to be difficult to manufacture with another feature expected to perform an analogous function to that of the removed feature. Four conceptual fuel element design candidates, with various fabrication features, for conversion of HFIR with 4.8 gU/cm3 LEU3Si2-Al fuel have been generated and shown to meet pre-defined performance and safety metrics. Results to date indicate that HFIR could convert with the subject fuel system and meet performance and safety requirements if, among other considerations, fabrication of the specific design features are demonstrated and qualification of the fuel is complete under HFIR-specific conditions.

Chandler, David↗
DOE OSTI2023

Sensitive Detection of an Erbium Isotope in an Atomic Beam Using Cavity Ring-Down Spectroscopy

Here, we have applied cavity ring-down spectroscopy (CRDS) to the study of the 166Er isotope in an atomic beam. These measurements were realized with an external cavity diode laser tuned to the 400.9 nm atomic transition of erbium and a customized high-finesse ring-down optical cavity under vacuum. Erbium atomic beams of different number densities were generated in a tantalum foil micro-crucible within the cavity. Absorbance values of the 166Er isotope between 3 × 10–6 and 7 × 10–5 were measured with a best-case precision on the order of 10–6, which is remarkable when considering the extreme temperatures at which the measurements were conducted, and the short detection path which is characteristic of collimated atomic beams. Number densities of erbium atoms were inferred to be between 2 × 106 and 6 × 107 cm–3. This work demonstrates for the first time the ability of studying dilute atomic beams of refractory materials with high accuracy utilizing CRDS. In these initial studies, we used erbium as a model system, but we expect to extend the proposed approach to the measurement of isotopes of uranium and plutonium for nuclear non-proliferation applications.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Dynamic separation of gases using microsieves

Separation of light weight molecules, such as nitrogen, argon, and oxygen, from heavier compounds can have significant impacts on energy capture, environmental monitoring, or isotopic applications. Large-scale gas separation techniques, like gas centrifugation and membrane mitigation, can be problematic as they impart tremendous energy and induce high mechanical stress onto the instrumentation. Microsieves, also known as micronozzles or microfunnels, are developed to create physical barriers to separate specific isotopes and gases. Separation is achieved using a converging and diverging micronozzle to impose supersonic gas flow around a curved wall, and it has been used for the separation of heavy actinide isotopes in low weight gas as well as separation of low weight gas compositions of nitrogen and argon back in 1900s. However, systematic reviews of this unique technology are lacking. The application of the Laval style nozzle, which has a converging/diverging entrance fundamental to the micronozzle, is included in this review due to its importance in industrial applications in uranium (U) isotope refinement. Using advanced computational fluid dynamic (CFD) simulations, the extent of gas separation can be modelled. Herein, we first examine the literature and survey recent advances on fabrication techniques for creating curved micronozzles, methods and separation principles used to design devices. Furthermore, we then follow with highlights of CFD simulations applied to evaluate the separation effects using microsieves. Finally, identification of the gap and recommendation for future development and applications are suggested for using intrinsic molecular features and fluidic dynamics in formulating separation strategies.

30 Microfluidics↗
DOE OSTI2022

The oxygen stable isotope composition of CRM 125-A UO 2 standard reference material

While there is a clear need for standardized reference materials for analytical calibrations and for inter-laboratory comparisons, there are not currently any for the oxygen stable isotopic composition of uranium oxides. In this paper we summarize the results from four laboratories by seven different methods of oxygen stable isotope analyses using fluorination techniques of CRM 125-A UO 2 Standard Reference Material. We synthesize these data and methods to arrive at a consensus oxygen stable isotope composition for CRM 125-A $δ$ 18 O = -9.63‰ (±0.29‰) VSMOW. We discuss methodological differences between analytical approaches, including furnace vs laser heating, fluorination using BrF 5 or ClF 3 , as well as calibration strategies. We highlight the potential effects of calibration scale compression from single-point calibrations using reference material with $δ$ 18 O values having a large relative difference from the sample being analyzed. We demonstrate how calibration scale compression can yield differences in calibrated $δ$ 18 O values up to ~2‰ for samples with ~20‰ difference from a single reference material, if the calibration slope of different analytical systems differs by 0.1. In conclusion, we suggest the use of liquid water calibration standards sealed in silver capillary tubes for multi-point calibrations of fluorination analysis systems.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2021

2021 Microbial U (Summary Report)

The US nuclear industry is on the cusp of significant growth in the areas of advanced reactors, SMRs, and microreactors. The existing commercial nuclear reactor fleet is considering expanding the life of the reactor cores by slightly increasing enrichment of their fuel. The common thread is the demand for High Assay Low Enriched Uranium (HALEU). A technical challenge exists with a blend down of Highly Enriched Uranium (HEU) from spent nuclear fuel (SNF) to meet the American Society for Testing and Materials (ASTM) and the “Y12” specifications. U-236 impacts reactor performance and is not desired in fresh fuel cores. In order to meet tight specifications, U-236 must be diluted by addition of fresh HEU. Fresh HEU is highly sought after and not plentiful in the US. A solution to this issue lies in the application of a technology to selectively remove the U-236 with non-hazardous microorganisms, which will be evaluated in this project by experimentally separating different isotopes of uranium.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Vadose and Saturated Zone Flow and Transport Calculations for the Active Trenches of the Low-Level Burial Grounds, Hanford Site, Washington

The purpose of the fate and transport modeling described in this environmental calculation file (ECF) is to evaluate the impacts to groundwater associated with waste disposal operations at Mixed Waste Low-Level Burial Ground (LLBG) Trenches 31 and 34 to satisfy requirements in DOE O 435.1, Radioactive Waste Management. The model integrates the flow and transport in the vadose zone beneath the active trenches with the saturated zone downgradient of the trenches to predict the radionuclide concentration at the point of assessment (POA). DOE M 435.1-1, Radioactive Waste Management Manual, defines the POA as the point of highest projected dose or concentration beyond a 100 m (328 ft) buffer zone surrounding the disposed waste. The modeling is conducted in accordance with the DOE G 435.1, Implementation Guide for Use with DOE M 435.1-1, performance assessment (PA) guidelines. The modeling involves evaluation of the groundwater concentrations and radionuclide arrival times during the 1,000-year compliance and 10,000-year sensitivity-uncertainty periods per DOE O 435.1 and DOE M 435.1. This analysis does not consider radionuclide release during facility operations, only the post-closure impacts of the radionuclides to the environment. The evaluation of potential radiological dose to groundwater receptors caused by releases from a closed facility containing radioactive waste typically includes the following: (1) Release of radionuclides from that facility (2) Transport of those radionuclides through the environment, and (3) Exposure to humans to environmental concentration levels of those radionuclides The fate and transport three-dimensional (3D) model analysis involves the post-closure impacts to the environment of the technetium-99, iodone-129, and uranium (all isotopes in the waste). The residual inventory estimates include several radionuclides, but technetium-99 is typically responsible for almost all of the beta-gamma dose equivalent associated with groundwater (water resources) protection per 40 CFR 141, “National Primary Drinking Water Regulations” (e.g., see the results in WCH-520, Performance Assessment of Environmental Restoration Disposal Facility, Hanford Site, Washington; hereinafter referred to as the ERDF PA), and iodone-129 can also be a significant dose contributor for some waste (e.g., RPP-RPT-59958, Performance Assessment for the Integrated Disposal Facility, Hanford Site, Washington; hereinafter referred to as the IDF PA). Uranium does not typically factor significantly into the impacts to groundwater, even during the 10,000-year sensitivity-uncertainty period, but always remains of interest as a contaminant. This ECF does not address vadose and saturated zone modeling for Trench 94 of the 200 East Area LLBG. Current information confirms the validity of the low corrosion rate of the naval reactor plant carbon steel (HY-80), and the even lower corrosion rate of the nickel-iron-chromium alloy reactor vessel (Inconel Alloy 600) presented in DOE/EIS-0259, Final Environmental Impact Statement on the Disposal of Decommissioned, Defueled Cruiser, Ohio Class, and Los Angeles Class Naval Reactor Plants. Based on these low corrosion rates, the time to breach the reactor vessel to allow release of radionuclides from the activated metal of the reactor vessel internal structure is at least 10,000 years. This time to breach precludes the need to evaluate the vadose and saturated zone transport of contaminants released from the reactor compartment disposal packages in the 200 East Area LLBGs PA (CP-63826, Waste Release Model Package Report for the Active Trenches of the Low-Level Burial Grounds, Hanford Site, Washington).

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

Proven Technologies for the Solidification of Complex Liquid Radioactive Waste (LRW): Global Case Studies of Applications and Disposal Options - 20424

Legacy radioactive waste streams from the Cold War still exist and newly generated waste streams from nuclear power plants and research institutes go untreated and expose environmental hazards at many nuclear sites. The nature of the waste is diverse, depending upon the source or the process from which it originated. The most problematic waste streams include complex liquids such as organic (tri-butyl-phosphate TBP) solutions contaminated with Pu and U isotopes, mixed sludge types, high acid radioactive waste, H-3 contaminated organic and aqueous streams, etc. Technological, environmental and economic challenges exist for the treatment and disposal of such waste streams. A proven technology that has been applied to LRW on a global basis provides one option as a low-cost solution to legacy streams and small volume, highly complex LRW frequently found during decommissioning at nuclear power plants and weapons sites. The engineered polymer technology from Nochar, USA, is capable of solidifying standard and highly complex LLW and ILW waste streams for interim or final storage, or for incineration. (authors)

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

Solvent Recovery and Management at the Savannah River Site H-Canyon Facility

NIOWAVE, Inc. is a domestic supplier of medical and industrial isotopes from uranium and radium. The Savannah River National laboratory (SRNL) is currently providing support to NIOWAVE, which plans to deploy a superconducting electron linear accelerator (LINAC) to fission uranium for Mo-99 production without the need for a nuclear reactor or HEU. The uranium from the Mo-99 production targets will be purified using a modified PUREX (Plutonium Uranium Reduction Extraction) solvent extraction process to recover the uranium in the product stream. The uranium will then be precipitated as an oxalate which is calcined to U 3 O 8 to fabricate pellets for new Mo-99 targets. In previous support provided to NIOWAVE, the SRNL demonstrated a solvent washing process to remove degradation products from the tributyl phosphate (TBP) solvent used in the modified PUREX process under development for uranium recovery. To supplement this technology demonstration, NIOWAVE requested the SRNL to provide summary information on the solvent recovery and management activities which are used at the Savannah River Site (SRS) H-Canyon facility. An existing reference document for the reprocessing of irradiated HEU fuels at the SRS was used as the primary reference for the solvent management activities; although, other reference documents were used to provide supplementary information. The information provided includes a brief summary of the solvent degradation issues which have been observed in the H-Canyon solvent extraction cycles and resulting process safety concerns. The solvent recovery processes for the three cycles of solvent extraction used in the H-Canyon were subsequently described including the process equipment which consists of the continuous and batch solvent washers, pumps, and tanks. A final section is provided on the monitoring and analysis of solvent quality based on the previous work performed at the SRNL for NIOWAVE and past research and development activities performed to support the solvent extraction processes in both the SRS F-Canyon and H-Canyon facilities.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

NWAL Validation Report for U Assay by Potentiometric Titration (Modified Davies and Gray Technique) and U and PU Assay by Isotope Dilution Mass Spectrometry

SRNL Sensing & Metrology, previously Analytical Laboratory, is a member of the IAEA Network of Analytical Laboratories for Nuclear Material Measurement that is qualified to provide Uranium and Plutonium isotopic analysis of nuclear materials by Thermal Ionization Mass Spectrometry. As part of the membership, SRNL participated in the 2019 Nuclear Material Round Robin to maintain the qualification to provide isotopic analysis. SRNL also participated with the additional intent to qualify Uranium Assay by Potentiometric Titration (Davies and Gray method) and Uranium and Plutonium Isotope Dilution by Thermal Ionization Mass Spectrometry as part of the International Atomic Energy Agency’s Network of Analytical Laboratories. The proficiency test exercise was conducted by the Office of Safeguards Analytical Services, Department of Safeguards.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗