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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 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Hopper flows of mixtures of spherical and rod‐like particles via the multisphere method

Abstract The discharge of granular media from hoppers plays a vital role in many solids handling operations. The dynamic flow of poly‐disperse nonspherical particles differs markedly from mono‐disperse spheres and highlights the need to rigorously account for particle shape effects. The multisphere discrete element method (MS‐DEM) was utilized here to examine the bulk packing fraction, discharge rate, collisional stress tensor, and order parameter of a bi‐disperse sphere‐rod mixture evacuating a hopper. A modified form of the Myers and Sellers correlation with variable shape factor is capable of replicating the MS‐DEM discharge data. The collisional stress tensor is observed to be highly symmetric for mono‐disperse spheres but becomes asymmetric when rod‐like particles are introduced. The order parameter within the hopper increases significantly during discharge and the rod‐like particles undergo shear induced alignment with their long dimensions orthogonal to the outflow plane.

Lattanzi, Aaron M.↗
DOE OSTI2015

Rod-by-rod detailed Monte Carlo method of Cofrentes NPP as a commercial core critical - 14482

The detailed rod-by-rod three dimensional modeling of large cores such as the commercial reactors is now possible. The simultaneous improvement in fuel depletion systems based on Monte Carlo codes allows using the criticality of first and reload cores as critical experiments, usually defined as commercial core criticals (CCC). The interest of such critical configurations for reload cores is evident for the validation of codes and cross-sections used for burn-up credit. Nevertheless other areas of the fuel management can benefit from this capabilities since the realistic rod-by-rod detailed modelling of fuel assemblies can provide valuable help in the optimization of reload patterns and in the licensing of other production-oriented calculation tools In this presentation the results obtained in reactivity and power distribution in the analysis of Cofrentes NPP first criticality and of a low-burnup xenon-equilibrium state-point and its comparison with in-core instrumentation measurements are presented. The plan for development of a local rod-by-rod Monte Carlo fuel depletion capability is presented as well. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Titanium and Aluminum Rod-Replacement Experiments in Fully-Reflected 6.90% Enriched UO{sub 2} Fuel Rod Lattices

The Seven Percent Critical Experiment (7uPCX) was designed to provide benchmark criticality and reactor physics data for water-moderated pin-fueled nuclear reactor cores. The enrichment of the fuel was chosen to explore the enrichment range above the current 5% ceiling for US commercial pressurized water reactors. The experiment was part of the US Department of Energy (DOE) Nuclear Energy Research Initiative Project 01-124 titled 'Reactor Physics and Criticality Benchmark Evaluations for Advanced Nuclear Fuel'. Three sets of benchmark experiments have been completed using the 7uPCX and documented as LEUCOMP-THERM-080, LEU-COMP-THERM-078, and LEUCOMP-THERM-096. The experiments for LEUCOMP-THERM-080 and LEU-COMP-THERM-078 were done in fully-reflected arrays with the number of fuel rods in the array as the approach parameter. The experiments for LEU-COMP-THERM-096 are similar to those in LEUCOMP-THERM-080 except that the arrays are partially-reflected - the arrays were larger than would be possible with full reflection and the approach-to-critical experiments were done with the depth of the water in the critical assembly as the approach parameter. The experiments described here are similar to the fully-reflected experiments in LEU-COMP-THERM-080 with some of the central fuel rods in the array replaced by titanium rods, aluminum rods, and/or water. The experiments are designed to provide criticality safety benchmarks with significant reactivity from titanium and equivalent configurations with aluminum experiment rods. These experiments are documented as LEU-COMP-THERM-097. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Full Rod Length Gamma Spectroscopy Measurements of LWR Used Fuel Rods

Twenty-five high-burnup pressurized water reactor used fuel rods were scanned in the Irradiated Fuels Examination Laboratory at Oak Ridge National Laboratory with a high-resolution gamma spectrometer. A planar high-purity germanium detector was used across an energy range from 60 keV to 3 MeV. Gamma spectra of individual fuel rods were collected in 4.5 mm steps along the entire length of each rod. Additional spectra were collected on select rods in rotational increments of 30 degrees to provide data on the distribution of fission products along the rod perimeter.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Measurements of control rod efficiency in RBMK critical assembly upon dropping of the rods

The efficiency of control rods in the RBMK critical assembly was measured in the case where one manual-control rod (MCR) is dropped from a steady critical state, and several other MCRs were additionally dropped after 44 s. The measured number of neutrons in the assembly during and after dropping of the rods was used to calculate the efficiency values of the rods by solution of the system of point kinetics equations. A series of methods of the initial data treatment for determination of the desired values of reactivity without the calculated corrections were used.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Parametric Evaluation of SiC/SiC Composite Cladding with UO 2 Fuel for LWR Applications: Fuel Rod Interactions and Impact of Nonuniform Power Profile in Fuel Rod

SiC/SiC composites are candidates for accident tolerant fuel cladding in light water reactors. In the extreme nuclear reactor environment, SiC-based fuel cladding will be exposed to neutron damage, significant heat flux, and a corrosive environment. Here, to ensure reliable and safe operation of accident tolerant fuel cladding concepts such as SiC-based materials, it is important to assess thermo-mechanical performance under in-reactor conditions including irradiation and realistic temperature distributions. The effect of non-uniform dimensional changes caused by neutron irradiation with spatially varying temperatures, along with the closing of the fuel–cladding gap, on the stress development in the cladding over the course of irradiation were evaluated. The effect of non-uniform circumferential power profile in the fuel rod on the mechanical performance of the cladding is also evaluated. These analyses have been performed using the BISON fuel performance modeling code and the commercial finite element analysis code Abaqus. A constitutive model is constructed and solved numerically to predict the stress distribution in the cladding under normal operating conditions. The dependence of dimensions and thermophysical properties on irradiation dose and temperature has been incorporated into the models. Initial scoping results from parametric analyses provide time varying stress distributions in the cladding as well as the interaction of fuel rod with the cladding under different conditions of initial fuel rod-cladding gap and linear heat rate. It is found that a non-uniform circumferential power profile in the fuel rod may cause significant lateral bowing in the cladding, and motivates further analysis and evaluation.

Accident tolerant fuel↗
DOE OSTI2023

Fuel performance evaluation of two high burnup PWR core designs during normal operation, control rod withdrawal, and control rod ejection scenarios

There is interest among utilities to extend the current, 18-month operating cycle to 24 months. Economically, this extension would require greater than 5 % enrichment and peak rod average discharge burnup levels above 62 GWd/MTU. A notable challenge of increasing enrichment is the resulting additional excess reactivity encountered during the early stages of fuel life. To accommodate, burnable absorbers beyond soluble boron are introduced into the fuel system. In high burnup fuels, the possibilities of cladding lift-off and fuel melting increase due, in part, to increased rod internal pressures and limited fuel thermal conductivity, respectively. This work collaboratively employs PARCS, RELAP5-3D, and BISON to compare the fuel performance of two high burnup fuel candidates with higher than 5 % enrichment. Here, the fuel performance parameters were compared to current NRC guidance. The results demonstrate an annular fuel design with homogenously blended gadolinium as a burnable absorber operates with greater safety margins during normal operation, allowing for additional operational flexibility. During normal operation, the core design utilizing Integral Fuel Burnable Absorber pins contained fuel pins which reached plenum pressures above 15.5 MPa by the end of the first fuel cycle and fuel pins experienced cladding hoop strains above 1 %. In the Gd core design, only two observed pins experienced plenum pressures above 15.5 MPa and no pins exceeded 1 % cladding hoop strain. During the control rod withdrawal scenario, plenum pressures for pins in both designs marginally exceeded system pressure, however neither experienced excessive hoop strain. The Gd core design experienced a maximum fuel temperature of 2418 K, which is significantly higher than the Integral Fuel Burnable Absorber design at 2157 K, but still within regulatory guidance. We predicted that the fuel in both could return to service after the CRW event. We also predicted that cladding would not fail during the Control Rod Ejection in either core design. Generally, the Integral Fuel Burnable Absorber core design performed with greater safety margin with regards to temperature during normal operation and the transient events. However, the Gd core design performed with greater safety margin regarding plenum pressure and hoop strain limits during normal operation and both transient events.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2016

Ultrasonic Nondestructive Evaluation of Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) During Large-Scale Load Testing and Rod Push-Out Testing

The Pultruded Rod Stitched Efficient Unitized Structure (PRSEUS) is a structural concept developed by the Boeing Company to address the complex structural design aspects associated with a pressurized hybrid wing body (HWB) aircraft configuration. The HWB has long been a focus of NASA's environmentally responsible aviation (ERA) project, following a building block approach to structures development, culminating with the testing of a nearly full-scale multi-bay box (MBB), representing a segment of the pressurized, non-circular fuselage portion of the HWB. PRSEUS is an integral structural concept wherein skins, frames, stringers and tear straps made of variable number of layers of dry warp-knit carbon-fiber stacks are stitched together, then resin-infused and cured in an out-of-autoclave process. The PRSEUS concept has the potential for reducing the weight and cost and increasing the structural efficiency of transport aircraft structures. A key feature of PRSEUS is the damage-arresting nature of the stitches, which enables the use of fail-safe design principles. During the load testing of the MBB, ultrasonic nondestructive evaluation (NDE) was used to monitor several sites of intentional barely-visible impact damage (BVID) as well as to survey the areas surrounding the failure cracks after final loading to catastrophic failure. The damage-arresting ability of PRSEUS was confirmed by the results of NDE. In parallel with the large-scale structural testing of the MBB, mechanical tests were conducted of the PRSEUS rod-to-overwrap bonds, as measured by pushing the rod axially from a short length of stringer.

Johnston, Patrick H.↗
DOE OSTI2022

Pellet cladding mechanical interaction as a potential failure mechanism during a control rod drop accident in a boiling water reactor

Boiling water reactors (BWRs) represent approximately one-third of the operating fleet in the United States, contributing significantly towards the global effort in reducing carbon emissions. Even though most of the operating fleet has been in operation for quite some time, continued advancements in new nuclear fuel (such as accident tolerant fuel) or operating regimes (such as power up-rates and higher burnup operation) necessitates similar advancements in modeling and simulation capabilities. Bison, a component of the Virtual Environment for Reactor Applications (VERA), is a high-fidelity fuel performance code able to explore the fuel performance of a wide variety of fuel types in one-, two-, and three-dimensions. Until recently, the code had not been used for analyses of BWRs. Modeling capabilities have been added for Gd-bearing UO{sub 2} and pure zirconium liners. New models have been added based upon the U.S. Nuclear Regulatory Commission (NRC) guide-lines for hydrogen pickup in Zircaloy-2 claddings and pellet-clad mechanical interaction (PCMI) failure during a reactivity insertion accident (RIA), known as a rod drop accident (CRDA) in BWRs. Implementing and/or improving Bison modeling capabilities extended its analytical reach to areas beyond its original intended purpose. This paper demonstrates one of these capabilities as a proof of concept. Recently developed models enable Bison to provide an alternate approach for cladding integrity determination in CRDA evaluations, which currently use bounding conservative estimates. As part of this demonstration, NRC guidance on hydrogen-pickup and PCMI failure models were utilized in this research. Even though more research is needed in establishing right inputs and process in this area, this paper demonstrates Bison's ability to determine cladding integrity in a CRDA evaluation. This first of a kind demonstration is a proof of concept in this area, which could potentially be extended to a number of other areas where a more accurate cladding integrity determination would be needed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

A combined experimental and computational study of water-gas shift reaction over rod-shaped Ce 0.75 M 0.25 O 2 (M=Ti, Zr, and Mn) supported Cu catalysts

Water-gas shift (WGS) reaction over a series of ceria-based mixed oxides supported Cu catalysts was investigated using a combined experimental and theoretical method. The mixed rod-shaped Ce 0.75 M 0.25 O 2 (M = Ti 4+ , Zr 4+ , Mn 4+ ) solid solutions, which majorly expose the (110) and (100) facets, are synthesized by hydrothermal method and used to prepare supported Cu catalysts. We found that the Cu/Ce0.75Ti 0.25 O 2 (Cu-CT) exhibits the highest CO conversion in the temperature range of 150-250 °C among all supported Cu catalysts. This is mainly attributed to (i) good dispersion of Cu; (ii) largest amount of moderate copper oxide; and (iii) strongest Cu-support interaction of Cu-CT. And compared to other mixed metals, periodic density functional theory calculations performed, this work further suggest that the introduction of Ti into CeO 2 not only promotes oxygen vacancy formation and CO adsorption, but also facilitates the carboxyl (COOH) formation at the interface of the Cu cluster and the support, which leads to the enhanced catalytic activity of the Cu-CT toward WGS reaction.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Nuclear fuel rods and heat pipes in a graphite moderator matrix for a micro-reactor, with the fuel rods having fuel pellets in a BeO sleeve

A reactor unit cell is disclosed including a graphite moderator structure, a heat pipe positioned in the graphite moderator structure, and a fuel assembly positioned in the graphite moderator structure. The fuel assembly comprises at least one fuel rod. Each fuel rod comprises a beryllium-oxide sleeve and nuclear fuel positioned in the beryllium-oxide sleeve.

Levinsky, Alex↗