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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Effects of Chlorine Capture and a Proposed Density Law on the Reactivity of Plutonium Solution Systems

During fissionable material processing, all normal and credible abnormal conditions must remain safely subcritical. Nuclear Criticality Safety (NCS) uses a number of methods to determine subcriticality, one of which is the use of neutron transport codes such as MCNP6. In order to create models for use with MCNP6, both the geometry and materials in fissionable material processes must be known, or assumptions must be made and quantified for the impact to bias. One of the systems with a significant amount of bias due to material modeling assumptions is in the area of aqueous plutonium processing. These solutions are typically plutonium nitrate solutions or plutonium chloride solutions, which are modeled as fictitious plutonium metal-water mixtures because little is known about the actual density of the solution and there is no current predictive capability approved for use at Los Alamos National Laboratory (LANL) for modeling them. This research is currently underway to fill the gap and develop an algorithm for use with MCNP6 to model the density of plutonium chloride solutions. The method is to be validated with experimental data for density, and also validated with critical experiments using MCNP6. Note that the Chlorine Worth Study (CWS) was performed in December 2021 to help bridge the gap in chlorine data for critical experiments, and is currently awaiting International Criticality Safety Benchmark Evaluation Project (ICSBEP) review. This study was performed by LANL at the National Criticality Experiments Research Center (NCERC) at the Nevada National Security Site (NNSS). Additional information regarding this experiment may currently be found in LA-UR- 22-29180. Additionally, the Chemistry-Actinide Analytical Chemistry (C-AAC) at LANL has performed a number of solution density measurements for PuCl 3 -HC 1 -H 2 O, allowing for such data be used to create a semi-empirical density via the Pitzer method. The published dataset for the measurements is documented in LA-UR-22-25454. This method has already been tested successfully for aqueous plutonium nitrate solutions in SCALE. Current solution density measurements exist of plutonium concentrations of 0-~142g/L, all at 2M HC1, at temperatures 20-40°C. Additional data was taken for HC1-corrected density values, which essentially mimics the data for a pure PuCl x -water solution. The calculations in this report aim to support the current research by demonstrating the difference in system reactivity for the current modeling method when compared to the new proposed modeling with a density law implementation, which is being written as a Python tool to be used with MCNP6.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2022

On the Uranium and Plutonium Nuclear Data Evaluations [Slides]

This presentation provides overviews of uraniums, 233 U and 235 U, plutonium research including motivations, current status in ENDF/B-VIII.0, the inclusion of sub-thermal data, and the inclusion of LANL ratio capture-to-fission data. Additionally, the preliminary fit of Mosby’s data as reported, fluctuating neutron multiplicities, uncertainty in evaluated libraries, and experimental effects are also presented. The presentation concludes by providing a summary of plans for the U and Pu evaluation work.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Multidimensional modeling of fuel-cladding friction in an LWR fuel rod

Solving finite element problems with friction often increases the level of difficulty to obtain properly converged solutions. This type of challenge becomes more salient when advanced, possibly multiscale, material models are employed to capture the thermomechanical behavior of fuel and cladding materials. Here the present work details our recent developments in a nuclear fuel performance finite element code for the systematic consideration of friction in nuclear reactor finite element simulations. We show the application of friction and its effects on the mechanics of light-water reactor rods accounting for various fuel constitutive modeling techniques, model dimensionalities, pellet assumed geometries, and power conditions. In particular, we focus on the fuel rod mechanical behavior as it relates to fuel constitutive models, sensitivity to the coefficient of friction, pellet states of stress, and rod elongation. We discuss the trade-offs between the various multidimensional modeling options and highlight the relevance of frictional effects in the prediction of the fuel rod deformation and interfacial stresses. To relate our modeling results with actual reactor operation, simulations including frictional effects are compared with fuel rod elongation experimental data and some challenges for carrying out a full validation of the axial mechanics are discussed.

42 ENGINEERING↗
DOE OSTI2020

Report to NCSP on 2008 DANCE measurements of 233 U($\eta$,$\gamma$)

Uranium-233 plays an important role in the Th-U fuel cycle, with substantial production in the cycle. This cycle has been proposed as an alternative to the U-Pu fuel cycle due to its reduced production of transuranic elements. An accurate measurement of the 233 U($\eta$,$\gamma$) cross section is required by the National Criticality Safety Program (NCSP) to complete the neutron-induced cross section data, where experimental capture cross section data are scarce and were measured decades ago. The most recent capture cross section data available in the literature were measured in 2007 at the n_TOF facility (CERN); in the 60s measurements were performed at Rensselaer Polytechnic Institute (RPI) and at LANL. Finally, as reported by ORNL, a new evaluation with a revised (renormalized) fission cross section is needed on 233 U. The challenge for this measurement lies in the difficulty of measuring the capture cross section data in the competing fission background, as the fission cross section is around one order of magnitude larger than the capture cross section for 233 U. The accuracy of a capture cross section measurement depends on discrimination between $\gamma$’s produced in capture and fission reactions, for which an experimental setup combining capture and fission detectors is needed. For the ($\eta$,$\gamma$) cross section measurement at LANSCE, this discrimination is achieved by combining the Detector for Advanced Neutron Capture Experiments (DANCE), to measure $\gamma$’s from capture reactions, with a Parallel Plate Avalanche Counter (PPAC) to tag the $\gamma$’s produced by fission. This method was successfully used to measure 235 U and 239 Pu capture cross sections. In these measurements, the neutron capture cross section was determined in a large fission background well above 100 keV. As part of the NCSP nuclear data effort, we have looked at past DANCE measurements on 233 U($\eta$,$\gamma$) and evaluated whether existing data is adequate to apply this technique.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

304L Can Crush Validation Studies

Accurate prediction of ductile behavior of structural alloys up to and including failure is essential in component or system failure assessment, which is necessary for nuclear weapons alteration and life extensions programs of Sandia National Laboratories. Modeling such behavior requires computational capabilities to robustly capture strong nonlinearities (geometric and material), rate- dependent and temperature-dependent properties, and ductile failure mechanisms. This study's objective is to validate numerical simulations of a high-deformation crush of a stainless steel can. The process consists of identifying a suitable can geometry and loading conditions, conducting the laboratory testing, developing a high-quality Sierra/SM simulation, and then drawing comparisons between model and measurement to assess the fitness of the simulation in regards to material model (plasticity), finite element model construction, and failure model. Following previous material model calibration, a J 2 plasticity model with a microstructural BCJ failure model is employed to model the test specimen made of 304L stainless steel. Simulated results are verified and validated through mesh and mass-scaling convergence studies, parameter sensitivity studies, and a comparison to experimental data. The converged mesh and degree of mass-scaling are the mesh discretization with 140,372 elements, and a mass scaling with a target time increment of 1.0e-6 seconds and time step scale factor of 0.5, respectively. Results from the coupled thermal-mechanical explicit dynamic analysis are comparable to the experimental data. Simulated global force vs displacement (F/D) response predicts key points such as yield, ultimate, and kinks of the experimental F/D response. Furthermore, the final deformed shape of the can and field data predicted from the analysis are similar to that of the deformed can, as measured by 3D optical CMM scans and DIC data from the experiment.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Nuclear safety Enhanced: A Deep dive into current and future RAVEN applications

As the horizon of nuclear energy expands with the advent of small modular reactors, IV generation reactors, and fusion reactors, there is a growing perspective that the licensing process could benefit from a more comprehensive approach. Moving beyond traditional deterministic and PRA analysis might pave the way for a novel safety analysis paradigm propelled by the increasing computational power at our disposal. This paper explores different methodologies that can improve the outcomes of nuclear safety analysis. These range from uncertainty quantification techniques, aimed at enhancing the precision of safety margins, to deploying dynamic event trees by driving system code simulations, capturing the potential evolutions of severe accidents. These methodologies introduce innovative dimensions to safety analysis, considering the consequences of postulated events and the dynamics of accident sequences. However, they also bring forth challenges, especially in managing the complexity and sheer volume of potential scenarios. The paper touches upon some strategies to counter these challenges, emphasizing the importance of adaptability and continuous evolution in the face of emerging nuclear safety concerns. Additionally, the paper sheds light on the need for advanced tools to apply these methodologies. Among these tools is RAVEN, an open-source software designed for parametric and probabilistic analyses. Its core components, including distribution, sampler, and reduced order model, enable various applications, from risk assessment and mitigation to dynamic learning and plant control logic simulations.

97 - MATHEMATICS AND COMPUTING↗
DOE OSTI2025

Implementing Fuel Cladding Bonding and Assessing its Impact on Axial Gas Communication

Fuel rods irradiated in light-water reactors to burnup values above 45 MWd/kgU are subject to the formation of a chemical and mechanical bond between the fuel and cladding upon gap closure. The formation of the bond subsequently inhibits the ability of fission gases released from the fuel to flow freely to the plenum of the rod. The flowing of gases within fuel rods is referred to as axial gas communication. During transients, such as loss of coolant accidents, the bond may influence cladding deformation prior to breaking. Upon bond breakage, gases are able to more freely communicate to the lower pressure regions of the rod. The impact of bonding on gas communication and the ballooning behavior of the cladding during a loss of coolant accident is of interest to the nuclear industry in support of burnup extension for the existing light-water reactor fleet. In this report, a model to capture the effects of fuel cladding bonding in the BISON fuel performance code is presented. The theory of the model along with implementation testing is provided. A summary of a previously developed axial gas communication is given to set the stage for how the two models may be coupled together. A full-length pressurized-water reactor fuel rod demonstration is highlighted to evaluate the impact of including bonding on axial gas communication calculations using a preliminary coupling methodology. An overview of the next steps regarding the modeling of bonding, axial gas communication, and a more tightly coupled framework is also provided.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Revised 207 Pb Thermal Capture Gamma-ray Spectrum Evaluation

This report details the revision of the JENDL-4.0 thermal capture gamma-ray spectrum evaluation for 207 Pb and the comparison of the current evaluations and the revised spectrum. The history of the available reaction evaluations is described and the deficiencies are identified. The modified spectrum has an average energy of 7.329 MeV and an average multiplicity of 1.005, leading to a total energy released value of 7.366 MeV, just 0.03% off from the Q-value of the reaction, 7.368 MeV. Based on comparison to the measurements and structure libraries considered, the revised evaluation is closer to reality and should provide more accurate coupled neutron-gamma ray calculations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Effective field theory analysis of 3 He–α scattering data

We treat low-energy 3 He–α elastic scattering in an effective field theory (EFT) that exploits the separation of scales in this reaction. We compute the amplitude up to next-to-next-to-leading order, developing a hierarchy of the effective-range parameters (ERPs) that contribute at various orders. We use the resulting formalism to analyse data for recent measurements at center-of-mass energies of 0.38–3.12 MeV using the scattering of nuclei in inverse kinematics (SONIK) gas target at TRIUMF as well as older data in this energy regime. We employ a likelihood function that incorporates the theoretical uncertainty due to truncation of the EFT and use Markov chain Monte Carlo sampling to obtain the resulting posterior probability distribution. We find that the inclusion of a small amount of data on the analysing power A y is crucial to determine the sign of the p-wave splitting in such an analysis. Here, the combination of A y and SONIK data constrains all ERPs up to O(p 4 ) in both s- and p-waves quite well. The asymptotic normalisation coefficients and s-wave scattering length are consistent with a recent EFT analysis of the capture reaction 3 He(α, γ) 7 Be.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Modern nuclear and astrophysical constraints of dense matter in a redefined chiral approach

Here, we explore the quantum chromodynamics (QCD) phase diagram’s complexities, including quark deconfinement transitions, liquid-gas phase changes, and critical points by using the chiral mean-field (CMF) model that is able to capture all these features. We introduce a vector meson field redefinition within the CMF framework, enabling precise adjustments of meson masses and coupling strengths related to vector meson interactions. Performing a new fit to the deconfinement potential, we are able to replicate recent lattice QCD results, low-energy nuclear physics properties, neutron star observational data, and key phase diagram features as per modern constraints. This approach enhances our understanding of vector mesons’ roles in mediating nuclear interactions and their impact on the equation of state, contributing to a more comprehensive understanding of the QCD phase diagram and its implications for nuclear and astrophysical phenomena.

QCD phase transitions↗
DOE OSTI2021

Precision measurement of relative γ-ray intensities from the decay of 61Cu

A discrepancy, well outside reported uncertainties, has been observed between the accepted and measured values of the intensity ratio of the two strongest γ-ray following 61 Cu β + decay. This discrepancy has significant impact since the nat Ni(d,x) 61 Cu reaction has historically been one of only a few IAEA recommendations for use as a deuteron flux monitor and a considerable number of published cross sections measured in ratio to that beam monitor cross section may depend on the choice of either the first or second strongest ray in those calculations. To determine the magnitude of this error most precisely, over a hundred separate measurements of the 283 keV to 656 keV γ-ray emission ratio were collected from seven experiments and a variety of detectors and detection geometries. A weighted average of all these measurements indicates an error in the value listed in the Nuclear Data Sheets of 11% in either the primary or second-highest intensity γ-ray of 61 Cu, potentially introducing an 11% error in 61 Cu production cross section measurements, cross sections using nickel activation as a deuteron beam current monitor, or in dose rates when 61 Cu is used in nuclear medicine. General agreement with the Data Sheets with ten other intensity ratios suggests the most probable error is in the secondary (656 keV) emission, which accordingly should be updated from 10.8% to 9.69%.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Probing the bandgap and effects of t-BN domains in h-BN neutron detectors

Thermal neutron detectors in a lateral scheme were fabricated from a 70μm thick freestanding B-10 enriched hexagonal BN (h- 10 BN). Two sets of channel peaks corresponding to the neutron capture by 10 B occurring in h- 10 BN comprising turbostratic domains (t- 10 BN) have been recognized in the nuclear reaction pulsed height spectrum, from which a bandgap of 5.5 eV was directly deduced for t- 10 BN. Improved device performance over the prior state-of-the-art implies that the transport properties in the lateral plane of t-BN domains are sufficiently good and their presence in h-BN is not a showstopper for the further advancement of h-BN detector technologies.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2024

LDRD poster - Identifying the speciation of salt-based actinides in the presence of contaminants

Molten salt reactors (MSRs) are among the most promising Generation IV nuclear reactors. The salts used in these reactors exhibit a high affinity for moisture and oxygen, and heating does not eliminate the oxides that are formed. Additionally, due to the high operating temperatures, reactor parts experience pronounced corrosion, and the resulting corrosion products integrate into the fuel salt system. It is hypothesized that the speciation of f-elements in molten salts would change in the presence of oxides and corrosion products. These changes in local structure can alter the rheological properties, which are critical to the operational and safety parameters of the reactors. Therefore, an in-depth understanding of how corrosion products influence lanthanide/actinide speciation is crucial. To test this hypothesis, lanthanides (as surrogates for actinides) and actinide chlorides were mixed with oxides (such as nickel oxide and lithium oxide) and/or corrosion products (e.g., nickel chloride) and reacted in molten salts. The resulting phases were analyzed using techniques such as powder X-ray diffraction, Raman spectroscopy, UV-Visible spectroscopy, and nuclear magnetic resonance spectroscopy. In the presence of oxides, due to the high oxophilicity of f-elements, these elements capture oxygen from the oxides, forming lanthanide/actinide oxide or oxychloride phases. Specifically, with cerium chloride, the introduction of lithium oxide to the formation of various final products depending on the oxide concentration. At lower oxide concentrations, cerium oxychloride (CeOCl) in the +3 oxidation state was observed, while an increase in the oxide concentration led to the formation of cerium dioxide (CeO2), where cerium is in the +4 oxidation state. In contrast, with nickel oxide, CeOCl formation was consistently observed, regardless of the oxide concentration. This project successfully determined that the speciation of f-elements change in the presence of oxides and different oxides in molten salts result in different final products. When the influence of nickel chloride was examined, changes in f-element speciation were not observed. However, it reacted with lithium chloride to generate a lithium-nickel-chloride (Li6NiCl8) phase, which was also observed when nickel oxide was reacted. Computational models were developed to simulate the phases anticipated under experimental conditions.

37 - INORGANIC, ORGANIC, PHYSICAL AND ANALYTICAL C↗
DOE OSTI2023

Material Effects on Electron-Capture Decay in Cryogenic Sensors

Several current searches for physics beyond the standard model are based on measuring the electron-capture (EC) decay of radionuclides implanted into cryogenic high-resolution sensors. The sensitivity of these experiments has already reached the level where systematic effects related to atomic state energy changes from the host material are a limiting factor. One example is a neutrino mass study based on the nuclear EC decay of 7 Be to 7 Li inside cryogenic Ta-based sensors. Here, to understand the material effects at the required level, we use density-functional theory to model the electronic structure of lithium atoms in different atomic environments of the polycrystalline Ta absorber film. The calculations reveal that the Li 1s binding energies can vary by more than 2 eV due to insertion at different lattice sites, at grain boundaries, in disordered Ta, and in the vicinity of various impurities. However, the total range of Li 1s shifts does not exceed 4 eV, even for extreme amorphous disorder. Furthermore, when investigating the effects on the Li 2s levels, we find broadening of more than 5 eV due to hybridization with the Ta band structure. Material effects are shown to contribute significantly to peak broadening in Ta-based sensors that are used to search for physics beyond the standard model in the EC decay of 7 Be, but they do not explain the full extent of observed broadening. Understanding these in-medium effects will be required for current- and future-generation experiments that observe low-energy radiation from the EC decay of implanted isotopes to evaluate potential limitations on the measurement sensitivity.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Simulations of Criticality Control Overpack Container Compaction at the Waste Isolation Pilot Plant

Criticality Control Overpack (CCO) containers are being considered for the disposal of defense-related nuclear waste at the Waste Isolation Pilot Plant (WIPP). At WIPP, these containers would be placed in underground disposal rooms, which will naturally close and compact the containers closer to one another over several centuries. This report details simulations to predict the final container configuration as an input to nuclear criticality assessments. Each container was discretely modeled, including the plywood and stainless steel pipe inside the 55-gallon drum, in order to capture its complex mechanical behavior. Although these high-fidelity simulations were computationally intensive, several different material models were considered in an attempt to reasonably bound the horizontal and vertical compaction percentages. When exceptionally strong materials were used for the containers, the horizontal and vertical closure respectively stabilized at 43:9 % and 93:7 %. At the other extreme, when the containers completely degraded and the clay seams between the salt layers were glued, the horizontal and vertical closure reached respective final values of 48:6 % and 100 %.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Assessment of Nuclear Data Needs for Advanced Reactor Demonstrations: Application to the Molten Chloride Reactor Experiment (MCRE)

A number of advanced reactor concepts are planned for near-term demonstrations including microreactors, larger demonstrations and space nuclear systems. These reactor concepts are based on a wide variety of reactor technologies, including sodium, gas, and salt cooling. An overlooked area in the development and ultimate startup of these reactors is addressing nuclear data needs that allow confident prediction of the criticality, safety requirements, and operation of the reactors. In this paper we try to address the problem of assessing nuclear data needs and possible remedies to reduce the existing uncertainties for advanced nuclear reactors. A methodology for defining these needs is described. The case of the Molten Chloride Reactor Experiment (MCRE) has been considered and the related investigation highlights the specific needs for reducing uncertainty on the 235U capture and 35Cl (n,p) reactions. Integral experiments relatively inexpensive are indicated as possible solution for significantly reduce the current associated uncertainties.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗