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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 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

FAST-1.1: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients (Developed under NQA-1-2017)

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: * heat conduction through the fuel and other materials * heat transfer from the cladding-to-coolant * cladding elastic and plastic deformation, including creep * fuel-cladding mechanical interaction * fission gas release from the fuel * rod internal pressure and void volume * cladding oxidation The code contains necessary material and coolant properties, as well as clad-to-coolant heat transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.1 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.1 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

FAST-1.2: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include: • heat conduction through the fuel and other materials • heat transfer from the cladding-to-coolant • cladding elastic and plastic deformation, including creep • fuel-cladding mechanical interaction • fission gas release from the fuel • rod internal pressure and void volume • cladding oxidation The code contains necessary material and coolant properties, as well as clad-to-coolant heat-transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.2 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.2 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Renewable low-Z wall for fusion reactors with built-in tritium recovery (Final Technical Report)

This project pursued development of a novel renewable plasma-facing wall technology for fusion reactors. The technology is based on a slurry which can be easily delivered by delivery tubes to the reactor wall. The slurry dries at the hot reactor wall into pebble rods which are extruded out into the hot plasma, where the pebbles break off and fall along the reactor wall and can be recovered by gravity and re-used. The falling pebbles carry away heat and tritium and also protect the wall against large scale erosion or redeposition of material. The research focused on carbon-based pebble rods and demonstrated that pebble rods could be produced from slurry with tolerable levels of outgassing on a reactor-relevant timescale (< 5 minutes). Steady-state handling of reactor relevant (up to 50 MW/m 2 ) normal-incidence heat loads was demonstrated. Pebble release velocities were found to be sufficiently small (< 1 m/s) to allow recovery below the vacuum chamber. Tunability of the pebble rod breaking rate was demonstrated by changing the fill fraction of the interpebble matrix which binds the pebbles together. This work could benefit the public by helping move forward the design of commercially viable fusion energy reactors. Designing a first wall for magnetic fusion reactors which can handle the huge heat loads present and also avoid buildup of tritium-containing deposits is extremely challenging and requires novel approaches like the one being investigated here.

36 MATERIALS SCIENCE↗
DOE OSTI2026

High-Burnup LOCA Burst Susceptibility BISON Analysis in PWRs and BWRs

Accurately assessing high-burnup fuel behavior during loss-of-coolant accidents (LOCAs) is essential for understanding fuel fragmentation, relocation, and dispersal (FFRD) risks across the US light-water reactor fleet. This work updates previous Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program multiphysics LOCA analyses for a pressurized water reactor (PWR) and a boiling water reactor (BWR) by incorporating recent model and material property advancements in the BISON fuel performance code, including a high-burnup structure (HBS) model, revised cladding burst criteria, and updated thermal–mechanical correlations. This update was needed to support ongoing industry initiatives and upcoming regulatory changes. Full-core, rod-resolved operating histories generated using Virtual Environment for Reactor Analysis (VERA) and system-level LOCA conditions obtained from TRACE were applied to statistically representative rod samples in BISON to evaluate burst behavior and FFRD susceptibility. These calculations used two cladding burst correlations and three fuel pulverization models so that the predictions of these models could be compared. The updated PWR simulations show markedly improved numerical stability as the number of crashed simulations decreased by 95% compared to the previous study, and hence higher confidence in results. The updated PWR simulations predicted cladding bursts exclusively among once-burned, high-power rods, with two different cladding burst models identifying the same burst-susceptible population. Resulting FFRD susceptibility estimates are significantly reduced compared with earlier studies, driven by cooler predicted fuel and plenum temperatures, lower hoop strains, and reduced fission gas release in the updated models. In contrast, none of the BWR rods were predicted to burst under either burst criterion, reaffirming minimal BWR FFRD susceptibility even with updated HBS and material models. Comparisons between the PWR and BWR end-of-cycle predictions are made. Comparison with prior work highlights significant shifts in PWR fuel performance metrics and confirmation of earlier BWR conclusions. Overall, the updated results underscore the importance of having high-resolution detailed modeling capability and continuously integrating evolving material models and physics into high-resolution multiphysics simulations. The unified assessment presented here strengthens confidence in predicting high-burnup LOCA behavior by improving agreement between different cladding burst correlations. These results also provide an improved foundation for future BISON model development, FFRD susceptibility calculations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

FAST-1.2.2: A Computer Code for Thermal-Mechanical Nuclear Fuel Analysis under Steady-state and Transients

Fuel Analysis under Steady-state and Transients (FAST) is the U.S. Nuclear Regulatory Commission (NRC)’s computer code that calculates the steady-state and transient response of nuclear reactor fuel rods during long-term in-reactor burnup, anticipated operational occurrences (AOOs), design basis accidents (DBAs), and dry storage conditions. The code calculates the temperature, pressure, and deformation of a fuel rod as functions of time-dependent fuel rod power and coolant boundary conditions. The phenomena modeled by the code include heat conduction through the fuel and other materials, heat transfer from the cladding-to-coolant, cladding elastic and plastic deformation (including creep), fuel-cladding mechanical interaction, fission gas release from the fuel, rod internal pressure, void volume, and cladding oxidation. The code contains necessary material and coolant properties, as well as clad-to-coolant heat transfer correlations, for normal operation through postulated accidents and AOOs for today’s U.S.-based light water reactor (LWR) fuel designs. FAST-1.2.2 also contains preliminary materials and models for new LWR fuel concepts, such as accident tolerant fuel (ATF), and non-LWR fuel concepts such as metallic fuels for sodium fast reactors (SFRs). FAST has been developed for use on Windows and Linux operating systems. This document describes FAST-1.2.2 and is one of a series of documents on the code; the other documents detail the material properties used by FAST as well as its integral assessment to experiments and commercial data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Damaged Fuel in the United States

Over the years of commercial nuclear power plant operations in the United States (U.S.), many fuel assemblies have lost the capability to perform all of their desired functions. These assemblies have lost the capability to be handled, stored, or transported to meet regulations resulting in them being classified as damaged fuel. The causes of these failed assemblies are diverse and plant-specific. The majority of these failed assemblies are contained in a damaged fuel can that will be used in conjunction with a storage and/or transportation system. The storage and transportation systems have a limit on how many slots can be filled with damaged fuel cans. A damaged fuel can is generally a stainless steel container that confines damaged SNF and is closed on its end by screened openings. These screened openings allow gaseous and liquid media to escape but minimize the dispersal of gross particulate material. Out of an abundance of caution a few reactors have loaded high burnup fuel into damaged fuel cans. Damaged fuel is not licensed for storage or transport in the U.S., because the regulations for storage and transport do not specify exactly how to classify damaged fuel. Instead the regulations license/certify packages that specify approved contents. Damaged fuel must be included in the approved contents to be a viable option. In many cases damaged SNF is encapsulated in a damaged fuel can to ensure it can confine gross fuel particles, debris, and or damaged assemblies to a known volume within a loaded cask. This damaged fuel can may then be utilized in the same way as an assembly in a storage and transportation design system. Some storage cask systems utilize top and bottom plugs to confine debris in damaged fuel. The most recent domestic documentation on damaged SNF was performed by the U.S. Energy Information Administration (EIA) and used data from U.S. reactors concluding in June 30th 2013 to produce Form GC-859, “Nuclear Fuel Data Survey”. Based on this form there were 136,821 boiling water reactor (BWR) spent nuclear fuel (SNF) assemblies and 104,647 pressurized water reactor (PWR) SNF assemblies for a total of 241,468 SNF assemblies in the U.S. Out of these 241,468 SNF assemblies, 4,521 assemblies were classified as failed. Some assemblies were also disassembled and the fuel rods or pieces of fuel rods were combined together to make a consolidated assembly. A consolidated assembly may include damaged fuel, or it could have been consolidated as part of a demonstration project. The GC-859 data includes 2,550 consolidated assemblies containing between 0 and 264 whole fuel rods. These consolidated assemblies could be placed in single assembly canisters and stored in the spent fuel pool. For dry storage and transportation, a single assembly canister is generally placed in a damaged fuel can. In addition to the consolidated assemblies, 2,391 uncanistered fuel rod pieces exist, which were removed from 494 assemblies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Linear delta systems, hexapod systems, and related methods

A linear delta system includes a support frame, rails mounted to the support frame, linear actuators, each linear actuator configured to translate along a longitudinal length of a respective rail, pairs of parallel rods each coupled to the linear actuators, a platform coupled to a longitudinal end of each of the pairs of parallel rods opposite the respective linear actuator, and an object coupled to the platform. Longitudinal axes of the rails are oriented parallel to each other and lie within a common plane or an uncommon plane. A method of forming a linear delta system includes mounting rails to a support frame, the rails having longitudinal axes that are parallel to each other and lying within a common plane, coupling a linear actuator to each of the rails, coupling a pair of parallel rods to each linear actuator, and coupling a platform to the pairs of parallel rods.

Crawford, Anthony L.↗
DOE OSTI2021

Load-decoupling attachment system

A load-decoupling attachment system is configured to secure a component to a primary structure. The load-decoupling system includes a fore end coupling bracket that is configured to attach to a fore end of the component. A first tie rod is coupled to the fore end coupling bracket. The first tie rod is configured to couple to a first portion of the primary structure. A second tie rod is coupled to the fore end coupling bracket. The second tie rod configured to couple to a second portion of the primary structure. A universal joint mount assembly is configured to couple to an aft end of the component and a third portion of the primary structure.

Weathers, Jeffrey W.↗
DOE OSTI2021

Linear delta systems with additional degrees of freedom and related methods

A linear delta system includes a frame, rails secured to the frame, linear actuators, each linear actuator coupled to a respective rail and configured to translate along a longitudinal length of the respective rail, pairs of parallel rods each operably coupled to a respective linear actuator, a platform coupled to the pairs of parallel rods, structure configured to movable couple an object to the platform; and at least one degree of freedom imparting assembly including a profiled rod extending in a direction parallel to the rails and a drive unit configured to rotate the profiled rod, wherein the at least one degree of freedom imparting assembly is configured to impart a degree of freedom to the object.

Crawford, Anthony L.↗
DOE OSTI2021

Atmospheric-pressure ionization and fragmentation of molecules for structural elucidation

A solution-cathode glow discharge mass spectrometry (SCGD-MS) apparatus comprises a SCGD source and a mass spectrometer. The SCGD source may comprise conductive rods, a power source, and a capillary. A method for ionizing an analyte comprises flowing an electrically conductive liquid onto a conductive rod, applying an electric potential to a second conductive rod such that a plasma discharge forms between the first conductive rod and the electrically conductive liquid to produce ions, and separating the ions in a mass spectrometer. The analyte may be a polypeptide that may be contacted with trypsin. The analyte may be a solid, liquid, gas, chemical complex, or ion in solution. The method may comprise sequencing the polypeptide.

Shelley, Jacob T.↗
DOE OSTI2023

Experiments to Measure the Effect of Tantalum on Critical Systems

Sandia National Laboratories (SNL) and Oak Ridge National Laboratory (ORNL) collaborated to develop a capability to test the epithermal/intermediate cross sections of materials at the SNL critical experiment facility using the Seven Percent Critical Experiment (7uPCX) fuel. As a result, a new set of critical experiments has been designed to target the epithermal cross sections of tantalum (Ta) and is scheduled to be performed at SNL in 2023. These critical experiments will be evaluated for inclusion in the 2024 edition of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. The focus of these critical experiments is to measure the effects of Ta on the critical array size. The critical array size will be determined by an approach-to-critical experiment with the number of fuel rods in the array as a free parameter. The core configurations are designed to optimize the reactivity worth of Ta and the overall percentage of Ta absorption rates in the epithermal/intermediate energy range (0.625 eV – 100 keV). The baseline core configuration includes 7uPCX fuel rods set at a triangular pitch of 1.016 cm and a central dry test region that utilizes a cadmium liner for filtering out thermal neutrons. The central test region has locations for 85 Ta rods set at a triangular pitch of 0.813 cm. The Ta reactivity worth for the case with 85 Ta rods is approximately 2.55%, with the percentage of Ta absorption rates within the intermediate energy range at nearly 90%.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

ECAR-6581 Rev 0 Hydraulic Assessment of Conceptual Design of Ring-type Spacer for MARVEL Reactor

The fuel rods in the nuclear reactor core can be deformed due to various reasons, such as thermal expansion, the effect of irradiation, hydrodynamic effects, etc. The deformation of the fuel rods can lead to contact between them, which can deteriorate the coolability of the nuclear fuel and increase the potential risk of failure. To prevent the fuel rods from coming into contact with each other, a ring-type spacer may be introduced in the MARVEL reactor. However, the ring spacers can result in an additional pressure drop, which could restrict the natural circulation operation of the MARVEL reactor. Therefore, this ECAR will examine the additional pressure drop caused by the introduction of the ring-type spacer by comparing with the bare rod condition.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2020

Effects of secondary void-initiating particles on the steady-state crack growth resistance of high-strength steel

In this investigation, we focus on the dependence of stable crack growth resistance, as measured by the short-rod fracture toughness test, on secondary void-initiating particles such as AlN and Ti-based grain-refining precipitates in a variety of high-strength steels with tempered martensitic microstructures. The model developed by Ritchie and Thompson is modified to illustrate the significant amounts of toughening that can result from the refinement of secondary particles. Analysis of the data suggests that material strength is a predominant factor in increasing the short-rod fracture toughness relative to linear-elastic measures of initiation fracture toughness, but the extent of toughening is limited by the size and number density of secondary particles in the microstructure. The variation in estimates of secondary microvoid initiation and growth strains with precipitate size reinforce the notion that primary fracture at non-metallic inclusions and secondary fracture at smaller particles occur as sequential processes with a degree of concurrence that is dependent on the state of precipitation in both particle dispersions. Toughening in this connection is maximized by increases in microvoid growth strain that result from decreases in the size and areal number density of secondary void-initiating particles. Finally, the occurrence of transient instabilities during crack extension in short-rod specimens is explained with a phenomenological model that relates crack growth stability to natural variations in the dispersion of secondary void-initiating particles in the microstructure.

42 ENGINEERING↗
DOE OSTI2022

Verification of the AGREE and Serpent for the steady state HTR-10 benchmark problems

This paper provides results from steady state HTR-10 benchmark calculations performed using AGREE and Serpent that are compared to experimental results as well as calculations performed by INET. The purpose of completing this benchmark is to validate AGREE and Serpent for the prediction of HTGR operation so that they may ultimately be used to support licensing and deployment efforts of advanced reactors. The benchmark consists of several problems ranging from control rod worth calculations to k{sub eff} calculations at various temperatures. Overall, AGREE and Serpent show good agreement with the reference solutions and are effectively able to predict HTGR operation for a variety of steady state cases. The largest difference from the reference was for the initial core single rod worth, which is possibly due to the larger core helium cavity causing inaccuracies in the diffusion calculation whereas the control rod worth determined by Serpent and AGREE is much closer to the reference result for the full core loading. Moreover, in every benchmark problem, increasing the number of energy groups in the cross sections results in improved agreement of the AGREE result with the reference, although it also causes an increase in computation time. For problems with experimental data available, the accuracy of results generated by Serpent and AGREE is comparable to results obtained by other benchmark participants. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Investigation of the OECD/NEA PWR MOX/UO{sub 2} core transient benchmark using a coupled whole core pin-by-pin route in WIMS

This paper investigates the new CAMELOT coupling route developed in WIMS, which uses prepared tabulated cross-sections combined with the flux solver MERLIN module and the integrated sub-channel thermal hydraulics solver ARTHUR module to solve for the coupled core state. This coupling route is applied to the OECD/NEA pressurized water reactor (PWR) core transient benchmark exercise, involving 3-D neutronics and thermal-hydraulics analysis of a full-sized PWR core featuring MOX fuel at various core states, as well as simulating a control rod ejection transient. Analysis of the results for each part of the benchmark demonstrate that the high-resolution pin-by-pin analysis used in WIMS closely matches the other participants in the benchmark, with differences of between 20-200 pcm for k-effective, 1-2.5% RMS difference for assembly-averaged powers, 2-8% difference in rod worth calculations, < 200 ppm for calculations of critical boron concentration, zero difference in delayed neutron fraction calculation and a very close match for power response to the rod ejection transient. This is the first application of the WIMS-CAMELOT approach for the transient analysis of a full-sized reactor and as such further improvements such as reducing computational cost through parallelization, optimisation and memory reduction, are currently in active development. This paper demonstrates the accuracy and flexibility of the CAMELOT coupling route, providing a straightforward and consistent user image that can be easily used for a variety of modelling problems involving coupled neutronics and thermal-hydraulics. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Temperature Dependent Early-Stage Oxidation Dynamics of Cu(100) Film with Faceted Holes

Fundamental understanding of surface oxidation dynamics is critical for rational corrosion protection and advanced manufacturing of nanostructured oxides. In situ environmental TEM (ETEM) provides high spatial (nano- to atomic- scale) and temporal (< 0.1 s) resolution to investigate the early-stage oxidation/corrosion dynamics of metals and alloys. Thin samples with facets are widely used to enable cross-sectional observation of the oxidation dynamics in ETEM. However, how different facet orientations oxidize under the same conditions, and how these facets change the oxidation process, has not been investigated before. Here, using in situ ETEM, we systematically compare the oxidation dynamics of Cu(001) thin films, with faceted holes exposing {100} and {110} facets at temperatures ranging from 250–600 °C under 0.03 Pa O 2 . Oxidation preference is observed to change, from Cu(110) facets at lower temperatures to Cu(100) facets at ~ 500 °C. Oxide growth mechanisms change from outward growth on Cu 2 O surfaces at low temperatures, to inward growth on Cu-Cu 2 O interfaces at high temperatures. At high temperatures (500–600 °C), a rod-like Cu 2 O morphology is observed, with side facets of ~ {024} and top facets of {100} on Cu(100). This differs from the square-shaped Cu 2 O exposing {110} facets formed on Cu(001) surfaces. Rod-like oxides exhibit directional growth along their lengths with linear growth rates, regardless of rod length and width. This suggests that O from Cu(001) surfaces, rather than Cu(100) facets, serves as an O source for oxide growth. These results show a direct comparison of oxidation at different orientations with temperature, underscoring the temperature dependence of oxidation preference. Our results also suggest future in situ ETEM experiments viewing oxidation corrosion cross-sectionally should be cautious when oxide size is comparable with sample thickness, as the oxidizing mechanism may change due to sample thickness.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Manufacturing ODS Steels from GARS Powders by Friction Consolidation and Extrusion

Previous research suggested that friction-based processing is a promising method for fabricating oxide dispersion-strengthened (ODS) steel. In this study, we combined friction consolidation and friction extrusion to successfully manufacture ODS steel rods using precursor powder made with gas atomization reaction synthesis. We examined the microstructure evolution from the initial powder to the final extruded rod, which revealed the dispersion process of Y. Additionally, by comparing the microstructures of three rods extruded at different temperatures, we showed that low-temperature friction extrusion effectively enhanced microstructure uniformity and prevented grain coarsening, leading to improved mechanical properties. Furthermore, our findings provide practical guidelines for adjusting processing parameters in the production of ODS steel using friction-based processing.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Machine learning-driven descriptions of protein dynamics at solid-liquid interfaces

This chapter has described how ML has enabled quantitative analysis of HS-AFM data to discover the physical phenomena governing protein dynamics and ordering at solid-liquid interfaces. The research detailed in this chapter modeled the rotation models of protein nanorods, the discovery of which would otherwise not be possible. By tracking the trajectories of individual protein rods from frame to frame, it was possible to model Brownian type motion and behaviors and Levy-flight dynamics that had not previously been shown. We also described the application of the Python package AtomAI, which has been developed specifically to analyze and extract physical phenomena, providing exemplar code for training an ensemble of deep neural networks to produce the semantic segmentation of AFM data and functions for encoding and decoding local environments. We last described a combinatorial approach to analyze very noisy data with a densely covered substrate where the emergence of order for the protein liquid crystals could be elucidated. By combining the methods from Case 1 and 2, it was possible to obtain the center of mass and angle for each rod in the images and track the assembly of the rods over time into a 2D liquid crystal array on the surface of mica.

protein dynamics, solid-liquid interfaces, atomic ↗