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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 1,243 records · Page 69
All sourcesNASA NTRSDOE OSTI
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 OSTI2020

Variation in plutonium dioxide sputter yields for 1–5 keV Ar+ ions

An oxide layer with a known thickness and chemistry was grown on delta stabilized Pu and sputtered with 1–5 keV Ar+ ions over a range of incident ion angle between 22° and 72°. From the time required to remove the oxide layer, sputter yields of PuO2 were calculated. The sputter yields appear to increase with higher Ar+ ion beam energy in the range of 1–5 keV at an incident sputter ion angle of 42° and were found to increase with a decreasing angle of incidence up to 62°. The degree of oxide reduction induced during the sputter process was found to vary with the incident sputter ion angle.

Donald, Scott B.↗
DOE OSTI2021

Underlying simplicity of 5f unoccupied electronic structure

Using a simple empirical model based upon the bremsstrahlung isochromat spectroscopy of elemental Th, it is possible to explain the recent high energy resolution fluorescence detection measurements of UF 4 (n = 2) and UCd 11 (n = 3) as well as the new inverse photoelectron spectroscopy of Pu 2 O 3 (n = 5), where n is the 5f occupation number. Furthermore, a critical issue in this analysis is the assumption that the Th 5f states are essentially empty, which will be confirmed both experimentally and computationally. Thus, for 5f systems, this simple model provides a unified and consistent picture of 5f unoccupied density of states in simple, localized systems, as the 5f occupation varies in the early part of the series, for n = 0, 2, 3, and 5.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Effects cascade debris and helium bubbles on the strength of aged plutonium

The radioactive decay of aging Pu is dominated by α-decay. This persistent α-decay produces crystalline defects in the form of dislocation loops and helium bubbles that evolve with time. Comparable defects are produced in other metallic alloys when subject to neutron irradiation, and these defects are known to modify the plastic deformation of irradiated materials. Models have been developed for these irradiated materials and validated against experimental confirmations of yield strength and the concomitant microstructural evolution. In this paper, we deploy those previously developed models and apply their mechanics to plutonium aging.

Chemical elements↗
DOE OSTI2025

Compatibility of molten plutonium with wrought and additively manufactured metal crucibles

Understanding plutonium’s interaction with metals is crucial for optimizing pyrochemical operations, nuclear fuel containment, and various actinide processing techniques. Traditionally, tantalum crucibles are employed for plutonium processing due to their high durability, excellent temperature stability, and low solubility in plutonium. However, tantalum faces challenges such as plutonium wetting and diffusion, making surface coatings particularly important for crucibles in pyrochemical applications to enhance corrosion resistance against plutonium. Tantalum is also expensive and difficult to machine, prompting the need for advanced manufacturing techniques to address these challenges. Here, in this work, we investigate the interaction of Pu with tantalum and titanium crucibles fabricated using both traditional machining methods and laser powder bed fusion (LPBF) additive manufacturing (AM). LPBF-AM is an advanced technique that allows for the creation of complex geometries from traditionally difficult-to-machine metals by using a high-powered laser to build parts. Previous studies of conventional manufactured tantalum have utilized oxidation and carburization of the surface to mitigate plutonium wetting; however, no studies of surface modified LPBF-AM material have been undertaken. These studies are crucial, given the typical differences in the grain structure between conventional and LPBF-AM materials. All crucibles underwent differential scanning calorimetry to confirm the melting of plutonium. Subsequently, the crucibles were sectioned and mounted in epoxy for microstructural analysis using optical microscopy and scanning electron microscopy. This investigation, comparing the performance of wrought vs AM metal crucibles, provides a basis for future tooling applications in actinide processing techniques and can address the challenges associated with traditional machining, particularly in pyrochemical applications.

Actinides↗
DOE OSTI2024

Development of spectroscopic tools for nuclear forensics signatures development

The Savannah River National Laboratory (SRNL) has been developing advanced spectroscopic tools for the characterization of plutonium-bearing compounds with the intent to develop material and process signatures for nuclear forensics. Plutonium in a production, refining, or finishing facility will exist in many forms including oxide precursors (PuF4, PuF3, Pu oxalate, etc.), oxide, and metal. The ability to identify plutonium in each of these chemical forms and determine their processing history is crucial for the development of spectroscopic signatures. This presentation will focus on our work to develop and apply spectroscopic tools at SRNL using doubled-walled cells (DWC) to characterize the thermal decomposition of oxalates, calcination chemistry, alpha-decay-induced chemistry, age dating since last calcination, and other signatures related to plutonium processing.

Villa-Aleman, Eliel↗
DOE OSTI2023

Actinide targets for the synthesis of superheavy nuclei

The use of heavy actinide targets, including 243 Am, 240,242,244 Pu, 245,248 Cm, 249 Bk, and 249 Cf, irradiated by intense heavy ion beams of 48 Ca has resulted in a significant expansion of the periodic table since 2000, including the discovery of five new heaviest elements and more than 50 new isotopes. These actinide materials can only be produced by intense neutron irradiation in very high flux reactors followed by chemical processing and purification in specialized hot cell facilities available in only a few locations worldwide. This paper reviews the reactor production of heavy actinides, the recovery and chemical separation of actinide materials, and the preparation of actinide targets for superheavy element experiments. The focus is on 248 Cm, 249 Bk, mixed 249–251 Cf, and 254 Es, including current availabilities and new production processes. The impacts of new facilities, including the Superheavy Element Factory at Dubna, accelerator and separator upgrades at RIKEN, and proposed upgrades to the High Flux Isotope Reactor at Oak Ridge are also described. Examples of recent superheavy element research are discussed as well as future opportunities for superheavy research using actinide targets.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2023

Ni/NiO Reference Electrode Potential Measurements in Molten CaCl 2 –CaO

A direct oxide reduction (DOR) process is used to reduce PuO 2 to Pu metal via Ca metal. Molten CaCl 2 is commonly used as the electrolyte for this reaction. As the DOR reaction proceeds, CaO accumulates in the salt and must be measured in real-time to know when the reaction has gone to completion. Open circuit potential (OCP) was measured in this report between a reference electrode consisting of Ni/NiO contained in a MgO tube and a tungsten working electrode. Two methods of CaO additions were performed over several experiments, and both methods correlated to a 0.11 to 0.23 V decrease in OCP up to 14.4 wt% CaO. CaO concentrations were separately measured via sampling and acid-base titration. A plot of OCP versus concentration of CaO fits the Nernst equation with the CaO powder additions up to 5.2 wt% and also at the higher end of the CaO pellet concentration range. While the MgO tubes were observed to be structurally stable in the molten salt, scanning electron microscopy and energy dispersive X-ray spectroscopy showed that Mg in the tubes had been partially replaced with Ca.

36 MATERIALS SCIENCE↗
DOE OSTI2024

ENDF/B-VIII.1: Photonuclear Sublibrary

The photonuclear sublibrary aims to describe nuclear reactions between incident gamma particles and different nuclei. For ENDF/B-VIII.1, we adopted around 200 evaluations from the IAEA Coordinated Research Project (CRP) on photonuclear data while, for a few select mission-critical materials, evaluations from ENDF/B-VIII.0 were maintained. Additionally, many fixes to many files were implemented. Finally, ENDF/B-VIII.1 includes a new photonuclear file, 242 Pu, which was adopted from JENDL-5.0.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Neutron Reaction Sublibrary

The neutrons sublibrary aims to describe nuclear reactions between incident neutron particles and different nuclei. For ENDF/B-VIII.1, many neutron files were re-evaluated or received major changes. A new 239Pu evaluation was jointly-produced by IAEA, LANL, LLNL and ORNL bringing important updates to fission neutron multiplicity, Prompt Fission Neutron Spectrum, resonance and fast regions. Around one third of the new neutron evaluations were performed as part of the INDEN collaboration, including 16,18 O, 19 F, 28,29,30 Si, 63,65 Cu, 50,51,52,53,54 Cr, 55 Mn, 54,56,57 Fe, 139 La, 233,235,238 U, 240,241 Pu. Important non-INDEN evaluations include 234,236 U, 206,207,208 Pb, 181 Ta, 88 Sr, 140,142 Ce, Pt and Dy isotopes, and many others. Also, dosimetry reactions from IRDFF-II were adopted for many materials.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

ENDF/B-VIII.1: Neutron Induced Fission Product Yields Sublibrary

The neutron-induced fission yields (NFY) sublibray aims to describe fission yields in neutron-induced reactions. For ENDF/B-VIII.1, there were no new fission yields complete evaluations. However, an important fix the 241 Pu NFY file was implemented, solving a long-standing anomalous discontinuity issue accidentally introduced in ENDF/B-VI.2, decades ago. Additionally, many erroneously large uncertainties, present in NFY from ENDF/B-VIII.0, were corrected.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Measuring Nd(III) Solution Concentration in the Presence of Interfering Er(III) and Cu(II) Ions: A Partial Least Squares Analysis of Ultraviolet–Visible Spectra

Optical spectroscopy is a powerful characterization tool with applications ranging from fundamental studies to real-time process monitoring. However, it can be difficult to apply to complex samples that contain interfering analytes which are common in processing streams. Multivariate (chemometric) analysis has been examined for providing selectivity and accuracy to the analysis of optical spectra and expanding its potential applications. Here we will discuss chemometric modeling with an in-depth comparison to more simplistic analysis approaches and outline how chemometric modeling works while exploring the limits on modeling accuracy. Understanding the limitations of the chemometric model can provide better analytical assessment regarding the accuracy and precision of the analytical result. This will be explored in the context of UV–Vis absorbance of neodymium (Nd 3+ ) in the presence of interferents, erbium (Er 3+ ) and copper (Cu 2+ ) under conditions simulating the liquid–liquid extraction approach used to recycle plutonium (Pu) and uranium (U) in used nuclear fuel worldwide. Finally, the selected chemometric model, partial least squares regression, accurately quantifies Nd 3+ with a low percentage error in the presence of interfering analytes and even under conditions that the training set does not describe.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2022

Nuclide inventory characterization of EBR-II MOX fuel test pins

Characterization of 3 mixed oxide (MOX) fuel pins from the Experimental Breeder Reactor II (EBR-II) SPA-2/-2B experiment program was performed in support of the Advanced Reactor Experiments for Sodium Fast Reactor Fuels (ARES) joint project between Idaho National Laboratory (INL) and the Japan Atomic Energy Agency (JAEA). The characterization efforts were performed to enable transient testing in the Transient Reactor Test (TREAT) facility at INL and will include post irradiation examination (PIE) and other analyses. MOX fuel pins UW02009, 02011, and 06024 were historically irradiated from 1989 to 1994 for a total of 822.42 effective full power days (EFPDs) and peak burnup of approximately 14.3 at.% (atomic percent of heavy metal atoms, ∼134.4 GWd/t). They had since been placed in storage until a future use was identified. Characterization of the fission products and activation of each fuel pin was necessary to enable removal from storage and subsequent irradiation testing or PIE activities. The initial fuel pin geometry and composition were established using historical documentation. Irradiation history details were utilized to compute irradiated and decayed fuel pin masses, atomic densities, and activities. A pellet-wise axial neutron flux was also computed to assist in evaluation of the axial distribution of {sup 235}U and {sup 239}Pu, per pellet, within each of the test fuel pins. The results computed herein supported experiment design for advanced high-burnup MOX fuel designs under slow transient overpower (MOXTOP) conditions, which was Phase I of the ARES project. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Impact of nuclear data covariance libraries on uncertainty quantification of sodium cooled fast reactor simulation

Uncertainty quantification in core modelling to assess reactor performance is a regular practice to identify design safety limits. Best-estimate plus uncertainty (BEPU) methods are commonly used to propagate uncertainties to core responses from input parameters, characterized by variance-covariance libraries. This study evaluates the impact of different covariance libraries on the uncertainty associated with core safety parameters and their propagation through reactor transient modelling for a metallic core (ABR-1000). The COMAC-1.0, COMMARA-2.0, and COMMARA-2.1 covariance libraries are used to propagate nuclear data uncertainties to neutron feedback coefficients using the Generalized Perturbation Theory, then through transient safety analyses to establish the confidence interval for safety performance of the reactor using stochastic sampling. The considered transients include Unprotected Transient Over Power and Unprotected Loss Of Flow. Most influential nuclide and reaction pairs in different variance-covariance libraries are identified and their impact on the uncertainty of safety parameters are evaluated. Significant differences were observed between uncertainties from COMAC compared to COMMARA libraries. Uncertainties from COMMARA libraries were consistently larger than COMAC for all cases. For structure, coolant density feedback coefficients and control rod worth, total nuclear data uncertainties were approximately 50% larger for COMMARA than COMAC. Differences were also observed in identification of top contributors of uncertainty in terms of reaction-cross sections. For instance, largest contribution to total uncertainty of radial expansion coefficient from COMMARA-2.0 and 2.1 originates from Na{sup 23} P1 Elastic cross section (0.81%) but from Pu{sup 239} Fission (0.53%) with COM C-1.0. Such differences are evaluated in this research for all quantities of interest. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Effect of Nuclear Data Covariances on Integral Experiment Design with Sensitivity and Uncertainty Analysis

Washington River Protection Solutions (WRPS) uses MCNP6.2 and the Whisper code for criticality safety analyses of the Hanford Tank Farm. Together the codes derive baseline upper subcritical limits (USLs) for the waste models using experimental benchmarks. Whisper returns higher USLs, i.e. , has less of a conservative penalty, when the neutronic similarity of the experimental benchmarks to the application is high. Unfortunately, few critical benchmarks have high similarity to the Hanford tanks. The waste in the tanks is highly dilute in plutonium and contains large masses of weakly neutron-absorbing elements like iron and manganese. Experimental benchmarks typically have low sensitivity to these absorbers because they are present as structural materials. Lacking similar benchmarks, new Thermal Epithermal eXperiment (TEX) configurations with high Pu content and interstitial iron absorbers have been designed for the criticality safety validation. The features of the design have been iterated upon to maximize the similarity between the experiment and different Hanford waste models. The similarity is quantified with sensitivity analysis and uncertainty quantification using the representativity coefficient, or c k . The representativity calculation requires nuclear data covariances, which may differ between nuclear data libraries and between library versions. Because of these variations, the optimal design may depend on the nuclear data covariances library. A scenario can be envisioned where an experiment is designed, and c k is maximized, with one set of covariance data. However, when the covariance data is changed, say from ENDF/B-VII.1 to ENDF/B-VIII.0, and the benchmark is used in a criticality safety evaluation, the experiment becomes suboptimal with respect to c k . In this paper, we present how the optimal design of the new TEX experiments varied depending on the nuclear data covariances used to calculate c k . We compare ENDF/B-VII.1 and ENDF/B-VIII.0, as if the library had been updated since the design of the experiment. Additionally, we use JEFF3.3 to simulate if the covariance data of a different library had been used. The results show that the covariances do have an important effect on the designs, less so for thermal systems (where the data are more consistent between evaluations) and more so for epithermal systems where more differences exist.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Discovery of AMPX Thermal Scattering Law Processing Issue for Solid Moderators

The 2020 edition of The International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook includes a newly produced plastic-moderated evaluation, identified as PU-MET-MIXED-002 and referred to as PMM002. The 2021 edition should include another new plastic-moderated experiment, HEU-MET-THERM-004, referred to as HMT-004. These evaluations are of particular interest, as PMM-002 is moderated with polyethylene, and HMT-004 is moderated with polymethyl methacrylate (Lucite), allowing for investigation of the differences in the thermal scattering laws (TSLs) available for these materials. ENDF/B-VIII.0 includes newly produced data for 1 H-based scattering materials, including Lucite and yttrium hydride. Other available 1 H TSLs include polyethylene, light water, zirconium hydride. The unbound 1 H (free-gas) cross sections were also used. ENDF/B-VIII.0 also includes revisions made to the ENDF/B-VII.1 light water and polyethylene evaluations. Zirconium hydride was unchanged from ENDF/B-VII.1. As of this writing, there are no benchmarks included in the Verified, Archived Library of Inputs and Data (VALID) that are primarily moderated with any solid moderator, so validation that includes the PMM-002 and HMT-004 evaluations would expand the coverage to two new moderators. A study was undertaken at Oak Ridge National Laboratory with two purposes: to provide validation data based on systems that are primarily moderated with polyethylene and Lucite, and to demonstrate the reactivity changes that can result from use of different 1 H TSLs in these plastic-moderated systems. The first objective was mainly to test data and code for SCALE and AMPX, whereas the second objective was to provide useful data for practitioners on the potential variability of predicted $k_{eff}$ based on TSL changes. The use of an exactly correct TSL is often not possible given the materials involved (e.g., lubricants), but this study was intended to provide some indication of the magnitude of the changes among similar materials that can manifest through application of different TSLs in hydrogenous systems over a range of different neutron energy spectra. Unfortunately, the nominal $k_{eff}$ results exposed deviations between SCALE and Monte Carlo N-Particle (MCNP). Some of the results using different TSLs were also unexpected and difficult to explain. An investigation revealed a processing issue in AMPX that was caused by an ambiguous description of the data for incoherent elastic scattering in the ENDF manual. This issue is discussed in detail in a SCALE User Notification, and it affects some solid moderators, but it does not apply to 1 H bound in water. This issue highlights the importance of validating all TSLs used in safety analysis calculations. The remainder of this paper provides a more detailed examination of the results that triggered the investigation and a summary of the findings of that investigation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Validation of Jezebel Reactivity Coefficients and Sensitivity Analysis

Nuclear data validation is often performed today using criticality measurements. The gold standard for criticality measurements is the International Criticality Safety Benchmark Experiment Project (ICSBEP). The validation specifically focuses on the effective multiplication factor ($k_{eff}$). $K_{eff}$ is a relatively easy parameter to infer and has reduced uncertainty due to being at or above critical. However, while $k_{eff}$ is the most documented parameter and its uncertainties and sensitivities have been evaluated in great detail, it cannot be used as a standalone metric to determine inaccuracies in nuclear data (e.g., cross section data, PFNS, nu), which is based on theory, physics, and differential measurements. The Experiments Underpinned by Computational Learning for Improvements in nuclear Data (EUCLID) project aims to identify compensating errors in specific isotope nuclear data by optimally designing experiments that are, or are not sensitive to a suite of measurement parameters beyond $k_{eff}$. By identifying parameters that are sensitive to each other, oppositely sensitive, or have substantial magnitude differences in sensitivity, experiments can be designed to constrain questionable nuclear data. One sensitivity that is of particular interest to this project includes the sensitivity of reactivity coefficients. Reactivity coefficients compare reactivity, which is related to $k_{eff}$ at two different states therefore being sensitive to small changes in the system. The most common type of reactivity coefficient measurements is comparison to void for a small sample within the assembly. It is key that the sample sizes are small enough to not affect the flux of the full system. Reactivity coefficients were evaluated for many early experiments to better understand transport corrected cross sections. In fact, ICSBEP includes reactivity coefficient results as “Supplemental Measurements” in appendices for a handful of older benchmarks. One of those benchmarks is Jezebel, the bare Pu critical assembly. This paper compares new simulations of reactivity coefficients for Jezebel, and explores the sensitivity of reactivity coefficients to small changes in nuclear data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Impact of Spontaneous Fission Neutron Emission Energy on Neutron Detector Response

A parametric study was simulated in MCNP to investigate the impact of spontaneous fission neutron energy on neutron detector response. A total of 880 configurations were simulated by varying 3 He pressure and polyethylene reflector thickness for 8 different isotopes. Results showed that deviations between the simulated isotopes become more prevalent at higher energies with 238 U showing the greatest deviation. It was also observed that configurations comparable to most detection setups, 10 atm and 6 - 8 cm polyethylene, were closer to the upper bound of deviations from a 252 Cf. 240 Pu, a common spontaneous fission source of interest, ranged from 2% - 3% within this space. This suggests that further investigation is required to determine exact correction factors when calibrating neutron detectors with a 252 Cf. Although demonstrated here for a 4π-detection system, future plans include simulating a more detailed model of the MC-15 detection setup to determine exact correction factors for future experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗