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At least 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Flow boiling in geometrically modified microchannels

Microchannels are a promising solution for high-heat-flux thermal management scenarios, including high-performance microelectronics cooling and power electronics cooling. However, thermohydraulic instabilities result from the rapid vapor bubble formation. The prior literature has examined several methods, including constricted inlet microchannels, expanding microchannels, and auxiliary jetting microchannels, to mitigate the effect of these instabilities. Computational fluid dynamics and heat transfer (CFD/HT) modeling of the flow boiling phenomena in these microchannel configurations has seen limited examination, and one-to-one numerical comparisons of the different mitigation strategies have not been performed. In the present investigation, CFD/HT analyses using a three-dimensional (3D) volume of fluid model coupled with a phase-change model for the interfacial heat and mass transfer were performed for multiple microchannel configurations (constricted inlet, expanding, and auxiliary jetting microchannels). A benchmark case of a rectangular microchannel was examined to quantify baseline thermohydraulic performance. Results demonstrated slight to significant thermal performance improvements for all cases, and significant pressure benefits for the expanding and jetting cases, consistent with experimental results in the literature. Bubble dynamics and visualization for the baseline and alternative configurations are provided to give insight into their underlying physics, and the differences in performance were investigated and compared with available literature.

Broughton, Justin (ORCID:000000026848457X)↗
DOE OSTI2021

Measurement of the onset of nucleate boiling in liquid xenon

We report the first precision measurement of the superheat temperature required for bubble nucleation in liquid xenon of Δ T wall,ONB = (16.9±0.5) K and Δ T wall,ONB = (19.2 +0.4 -1.1 ) K at pressures of P = (0.98±0.02) bar and P = (1.32 +0.05 -0.01 ) bar, respectively. Both are measured at a subcooled bulk fluid temperature of ~ 162 K. A dedicated liquid xenon setup is used to measure the temperature at which bubble nucleation first appears on the flat surface of a resistor. The associated average heat flux at the surface of the resistor is determined using fluid dynamics simulations. Future experiments can use these experimental results to determine the likelihood of boiling in liquid xenon and to design effective thermal sinks that prevent bubble nucleation.

47 OTHER INSTRUMENTATION↗
DOE OSTI2019

Impact of Control Blade Insertion on the Deformation Behavior of SiC-SiC Channel Boxes in Boiling Water Reactors

This report describes the analysis of distortion of a Silicon carbide fiber-reinforced silicon carbide matrix (SiC-SiC) composite channel box under in-reactor conditions of a boiling water reactor (BWR). The BWR core has significant gradients in the fast neutron flux across the channel box due to the presence of water rods within the fuel assemblies, and these gradients increase further with the insertion of control blades. As a result of temperature and neutron flux dependent irradiation induced swelling of SiC, the SiC-SiC composite channel box can undergo significant distortion. In this work, we evaluate the SiC-SiC channel box distortion for three different control blade positions. This analysis is based on the neutron flux and temperature distributions in the BWR core calculated using the neutronics code MPACT and thermalhydraulics code CTF. This calculation is coupled through temperature feedback. Subsequently, we have performed structural analysis based on the calculated neutron flux and temperature distributions, to determine the deformation and stress development in the channel box. The structural analysis was performed using the fuel performance modeling code BISON and the commercial finite element analysis software Abaqus. The results indicate that the channel box will undergo time dependent lateral bowing for all the control blade positions in the assembly. The bowing behavior is dominated by the swelling of SiC-SiC material under non-uniform neutron flux, and changes with variation in the control blade position. The lateral bending will cause a temporary interference between the channel box and the control blade. The developed stresses exceed the proportional limit stress of the material, which may cause matrix microcracking in the channel box. However, the stresses remain below the ultimate tensile strength of the material, and therefore, development of a full, through thickness crack in the channel box is not expected.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Source Term Reduction for Advanced and Small Modular Boiling Water Reactors

The United States Department of Energy (US DOE) is currently supporting the development of various advanced and small modular reactor (SMR) designs. Several of these designs have commenced license application with the US Nuclear Regulatory Commission (US NRC). These reactors have improved safety features that may significantly reduce radiological source terms in the event of design and beyond-design basis accidents. Specifically, some reactors feature a smaller containment volume relative to the available fission product depositional surface area, which supports increased fission product retention in the containment vessel. Pressurized water reactors (PWR) and boiling water reactors (BWR) with this feature include the integral pressurized water reactor (iPWR) and the BWRX-300 design by General Electric. A prior research program supported by the US DOE quantified the source term reduction associated with light water iPWRs and developed iPWR-specific theoretical models for fission product deposition rates. This program included a sequence of research projects that started with a feasibility study, development of theoretical models that predict higher deposition rates, and finally, development of empirical data for verification and validation of the theoretical models. The current project, which is a feasibility study, is the first step in a similar program to quantify the source term reduction associated with small and advanced light water BWRs.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Report Summarizing Boiling Transition (Dryout) Testing of FeCrAl Cladding

Nuclear fuel and cladding in a light water reactor must maintain integrity under all normal operational and anticipated operational occurrences. In a Boiling Water Reactor (BWR) some Anticipated Operational Occurrences (AOOs) and Anticipated Transients Without SCRAM (ATWS) events have been known to result in cladding damage due to dryout with conventional Zircaloy-2 claddings. This work presents an assessment of commercially-fabricated, accident tolerant iron-chromium-aluminum (FeCrAl) alloy C26M cladding in such events with regard to the ballooning failure mode for unirradiated cladding In the severe accident test station (SATS), an experiment has been developed to gauge accident tolerance of fuel cladding materials under simulated cyclic dryout conditions. Thermal cycling from 300°C to 650°C was implemented on C26M and Zircaloy-2 tubes. The C26M tube did not fail after 54 cycles. The Zircaloy- 2 tube run under the same parameters as the C26M tube failed during the fourth cycle, consistent with expectations from the literature. Due to thermal gradients present in the furnace, both tubes experienced a maximum temperature at ~780°C. Due to differences in tube geometry, the hoop stress of the C26M tube was significantly higher than that of the Zircaloy-2 tube, 76 MPa at the maximum peak temperature compared to 55 MPa at burst, respectively. The cyclic dryout experiments show an increased performance of C26M over Zircaloy-2 under simulated conditions, an indication of C26M possessing superior high temperature oxidation resistance, mechanical properties and potentially resistance to fatigue.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Improvements to CTF for Modeling of Boiling Water Reactor Geometry and Operating Conditions

This report documents work done to improve CTF modeling of boiling water reactor (BWR) designs during FY 2020. Several improvements were made to the geometry modeling capabilities, including adding support for modeling of inlet flow orifices, axially varying large water rods, channel boxes with large corner round radii, and quarter symmetry models. Work was also done to improve on the numerical stability and run-time performance of CTF in Virtual Environment for Reactor Applications (VERA) for core-scale BWR problems. An outer iteration loop was implemented that ensures mass and energy are conserved before moving on to the next time step in the solution. The ability to use a direct solver for the pressure matrix was also added, which reduces the matrix solution error and improves code stability and run-time performance. Improvements to the pressure balance iteration loop were needed to allow CTF to be used in a coupled VERA solution of a model of Hatch that was developed. Finally, various improvements have led to a 2×reduction in the total VERA walltime for the Hatch model.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Summary of CTF Modeling and Numerical Improvements for Boiling Water Reactor Simulation

This report documents geometry and numerical improvements made to CTF for the modeling of boiling water reactor (BWR) geometry and operating conditions. These activities are part of a larger program to extend the Virtual Environment for Reactor Applications (VERA) to better support BWR modeling and simulation. The activities documented in this report added features to CTF, including support for mixed-fuel cores, modeling of the upper plenum, and modeling of the lower tie plate form losses. A review of the spacer grid modeling approach was also performed, and a plan was discussed for future improvement. The parallelization of the model was improved, leading to a roughly 2× improvement in CTF runtime and a 1.6× improvement in total VERA runtime. An in-depth review of the governing equations and their linearization was performed and is documented in this report. Once implemented, this new linearization will allow CTF to take much larger timesteps, leading to more significant reductions in CTF and VERA runtimes.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Boiling-Water Reactor Testing Capability in the Advanced Test Reactor

I-Loop is an irradiation facility that is currently being installed at the Advanced Test Reactor. It is a two-loop test facility capable of performing Light Water Reactor (LWR) irradiations in prototypic coolant conditions. The two loops are being installed to be capable of both Boiling Water Reactor (BWR) and Pressurized Water Reactor (PWR) pressure, temperature, and chemistry environments. Each loop is nominally dedicated as a BWR or PWR for simplicity of operations. In-reactor water loop testing that an I-Loop provides is key to the deployment of new accident tolerant fuel technologies and other advanced LWR fuel concepts. Currently, pressurized water loops are the only testing facilities available to test BWR fuel concepts. Their test environments are non-prototypic at higher pressure/temperature and at single-phase fluid flow conditions. This void in the LWR test bed capabilities is one that the I-Loop is uniquely situated to provide. This report discusses the mechanical design, thermal hydraulic calculations, and neutronic calculations of a proposed standard experiment of accident tolerant BWR fuel concepts. Mechanical design examines the geometry and features of the main components. Thermal hydraulic calculations examine the modeling and results of the two-phase flow options available. Lastly, neutronic calculations examine the Monte Carlo analysis of enrichment, heat rates, and flux spectrum.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Core Physics Characteristics of Extended Enrichment and High Burnup Boiling Water Reactor Fuel

This paper presents the highlights of boiling water reactor (BWR) core physics studies performed at Oak Ridge National Laboratory as part of a series of studies conducted to compare low-enriched uranium (LEU) with LEU+ fuel. The studies analyzed isotopic fuel content, lattice parameters (Phase 1), and core physics (Phase 2) to identify challenges in operation, storage, and transportation for BWRs and pressurized water reactors (PWRs). Because of a lack of publicly available lattice and core designs for modern BWR fuel assemblies and reactor cores, several optimized lattice designs were generated, and different core loading strategies were investigated. Twelve optimized lattice designs with 235 U enrichments ranging from 1.6% to 9% and gadolinia loadings ranging from 3 to 8 wt% were used to model axial enrichment and geometry variations in fuel assemblies for core designs. Each core shares a common set of approximations in design and analysis to allow for consistent comparisons between LEU and LEU+ fuel. The objective is to highlight anticipated changes in core behavior with respect to the reference LEU core. The results of this study show that the differences in LEU and LEU+ core reactor physics characteristics are less significant than the differences in lattice physics characteristics reported in the Phase 1 studies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Atmospheres of low-mass planets: The “boil-off”

We show that, for a low-mass planet that orbits its host star within a few tenths of an AU (like the majority of the Kepler planets), the atmosphere it was able to accumulate while embedded in the protoplanetary disk may not survive unscathed after the disk disperses. This gas envelope, if more massive than a few percent of the core (with a mass below 10M{sub ⊕}), has a cooling time that is much longer than the timescale on which the planet exits the disk. As such, it could not have contracted significantly from its original size, of the order of the Bondi radius. So a newly exposed protoplanet would be losing mass via a Parker wind that is catalyzed by the stellar continuum radiation. This represents an intermediate stage of mass-loss, occurring soon after the disk has dispersed, but before the EUV/X-ray driven photoevaporation becomes relevant. The surface mass-loss induces a mass movement within the envelope that advects internal heat outward. As a result, the planet atmosphere rapidly cools down and contracts, until it has reached a radius of the order of 0.1 Bondi radius, at which time the mass-loss effectively shuts down. Within a million years after the disk disperses, we find a planet that has only about 10% of its original envelope, and a Kelvin–Helmholtz time that is much longer than its actual age. We suggest that this “boil-off” process may be partially responsible for the lack of planets above a radius of 2.5R{sub ⊕} in the Kepler data, provided planet formation results in initial envelope masses of tens of percent.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2018

Hydrothermal Corrosion of Coatings on Silicon Carbide in Boiling Water Reactor Conditions

SiC/SiC ceramic matrix composites are an attractive material for use as accident tolerant fuel cladding. However, despite its resistance to corrosion in high-temperature steam, SiC is known to dissolve in high-temperature water, necessitating the use of corrosion-resistant coatings. Here, samples were exposed in 288°C water with 2 ppmw dissolved oxygen to simulate boiling water reactor (BWR) normal water chemistry (NWC) and 288°C water with 150 ppbw dissolved hydrogen to simulate BWR hydrogen water chemistry. Several candidate coatings were tested: Cr, CrN, TiN, and ZrN. All coatings were applied by physical vapor deposition onto chemical vapor deposition SiC (CVD-SiC) substrates. Two uncoated SiC/SiC variants were also exposed and degraded more rapidly compared to CVD-SiC, especially in NWC. Cr and CrN coated specimens were found to have low mass change, suggesting they are promising candidate coatings. Lastly, TiN coatings were found to oxidize rapidly, suggesting that coatings would fully oxidize during service in BWR environments. ZrN coatings dissolved completely during exposure.

36 MATERIALS SCIENCE↗