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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 1,423 records · Page 79
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Superluminal antenna

A superluminal antenna element integrates a balun element to better impedance match an input cable or waveguide to a dielectric radiator element, thus preventing stray reflections and consequent undesirable radiation. For example, a dielectric housing material can be used that has a cutout area. A cable can extend into the cutout area. A triangular conductor can function as an impedance transition. An additional cylindrical element functions as a sleeve balun to better impedance match the radiator element to the cable.

42 ENGINEERING↗
DOE OSTI2016

DNA reorganization in a mouse cell's nucleus

This visualization shows tightly-packed DNA in a mouse cell's nucleus at different stages of development, from its round shape as a multipotent stem cell; to a more oval shape as it develops into a more specialized neuronal progenitor; then a more triangular form as a mature nerve cell. The final, disjointed structure in this computerized rendering shows how removing a specialized binding protein (HP1β knockout) affects the structure of the DNA-packing material, called heterochromatin. (Credit: Berkeley Lab, UCSF)

Berkeley Lab, LBNL↗
DOE OSTI2015

Evaluation of a Potassium Filled UO2 Core in a Beryllium- ??eflected Space Reactor

The critical configuration of the small, compact critical assembly (SCCA) experiments performed at the Oak Ridge Critical Experiments Facility (ORCEF) in 1962-1965[1,2] have been evaluated for the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and were determined to be acceptable benchmark experiments and are included in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook)[3]. The initial intent of these experiments was to support the design of the Medium Power Reactor Experiment (MPRE) program, whose purpose was to study “power plants for the production of electrical power in space vehicles.”[4] The third configuration in this series of experiments was a beryllium-reflected mockup of a stainless-steel-clad, highly enriched uranium (HEU)-O2 fuel, potassium-cooled, space power reactor. Reactivity measurements, cadmium ratio spectral measurements, and fission rate measurements were measured through the core and top reflector. Fuel effect worth measurements and neutron moderating and absorbing material worths were also measured in the assembly fuel region. The cadmium ratios, fission rate, and worth measurements were evaluated for the International Reactor Physics Experiment Evaluation Project (IRPhEP) and are published in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)[5]. More recently, two additional critical configurations have been evaluated and added to the existing evaluation report in the ICSBEP Handbook. The new critical configurations measure the effect of filling the core with potassium. The critical configuration consisted of 30.48 cm tall fuel tubes of 93.15 wt.% enriched uranium dioxide (UO2) pellets. Each fuel tube had 26 pellets with a total mass of 295.8 g UO2 per tube. There were 253 tubes were arranged with 1.506-cm center-to-center, triangular spacing and surrounded on all side by a beryllium reflected. The reactivity of the system was measured with the core tank empty and filled with potassium. The system reactivity was increased from 13.4 ¢ to 32 ¢. This change in system reactivity represents the worth of the potassium. Simple benchmark models for the empty and potassium filled cores have been derived and evaluated and found to be acceptable as benchmark experiments. The benchmark experiment eigenvalues are 1.0001 ± 0.0008 and 1.0011 ± 0.0009 for the empty and filled cores, respectively. When calculated with MCNP5 and ENDF/B-VII.0 the calculated eigenvalues are 0.98906 ± 0.00002 and 0.98883 ± 0.00002, respectively. These results are concerning due to the drop of the calculated system reactivity when potassium is added to the system rather than the expected increase in system reactivity.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Subwavelength photonic crystal waveguide in optical systems

An optical system is disclosed. The optical system comprising: a substrate; and a subwavelength photonic crystal waveguide atop the substrate, wherein the subwavelength photonic crystal waveguide comprises a periodic one or two-dimensional array of two or more interleaved dielectric pillars; wherein the periodicity of the one or two-dimensional array is constant, a combination of two or more periods, or random; wherein the one or two-dimensional array is substantially linear or curved; wherein each of the pillars of the one or two-dimensional array is at least one of a triangular prism, a trapezoidal prism, an elliptic cylinder, a cylinder, a tube, a frustum, a pyramid, a trapezoidal prism, and an asymmetric frustum; and wherein each of the pillars of the one or two-dimensional array comprises a solid, liquid, and/or gas. Other embodiments are described and claimed.

Xu, Xiaochuan↗
DOE OSTI2020

Fabrication processes for effectively transparent contacts

In conventional solar cells with metal contacts, a non-negligible fraction of the incoming solar power is immediately lost either through absorption or reflection upon interaction with the contacts. Effectively transparent contacts (“ETCs”) for solar cells can be referred to as three-dimensional contacts designed to redirect incoming light onto a photoabsorbing surface of a solar cell. In many embodiments, the ETCs have triangular cross-sections. Such ETCs can be placed on a photoabsorbing surface such that at least one of their sides forms an angle with the photoabsorbing surface. In this configuration, the ETCs can redirect incident light onto the photoabsorbing surface, mitigating or eliminating reflection loss compared to conventional solar cells. When constructed in accordance with a number of embodiments of the invention, ETCs can be effectively transparent and highly conductive.

14 SOLAR ENERGY↗
DOE OSTI2016

DNA reorganization in a mouse cell's nucleus

This visualization shows tightly-packed DNA in a mouse cell's nucleus at different stages of development, from its round shape as a multipotent stem cell; to a more oval shape as it develops into a more specialized neuronal progenitor; then a more triangular form as a mature nerve cell. The final, disjointed structure in this computerized rendering shows how removing a specialized binding protein (HP1β knockout) affects the structure of the DNA-packing material, called heterochromatin. (Credit: Berkeley Lab, UCSF)

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2022

Graph Partitioning and Sparse Matrix Ordering using Reinforcement Learning and Graph Neural Networks

We present a novel method for graph partitioning, based on reinforcement learning and graph convolutional neural networks. Our approach is to recursively partition coarser representations of a given graph. The neural network is implemented using SAGE graph convolution layers, and trained using an advantage actor critic (A2C) agent. We present two variants, one for finding an edge separator that minimizes the normalized cut or quotient cut, and one that finds a small vertex separator. The vertex separators are then used to construct a nested dissection ordering to permute a sparse matrix so that its triangular factorization will incur less fill-in. The partitioning quality is compared with partitions obtained using METIS and SCOTCH, and the nested dissection ordering is evaluated in the sparse solver SuperLU. Our results show that the proposed method achieves similar partitioning quality as METIS and SCOTCH. Furthermore, the method generalizes across different classes of graphs, and works well on a variety of graphs from the SuiteSparse sparse matrix collection.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Molybdenum Sleeves Experiments in the Sandia Critical Experiments Facility

Sandia National Laboratories and the Institut de Radioprotection et de Sûreté Nucléaire have collaborated on the design and execution of a set of critical experiments that explore the effects of molybdenum in water-moderated fuel-rod arrays. The molybdenum was included as sleeves on some of the fuel rods in the critical experiment fuel arrays. Approach-to-critical experiments were performed on five configurations of fuel and molybdenum sleeves using the 7uPCX fuel in core hardware that set the triangular fuel rod pitch at 15.494 mm. The experiments are evaluated as benchmark critical experiments for the 2023 edition of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook as LEU-COMP-THERM-111.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

A multi-modal scanning system to digitize CBRNE emergency response scenes

A handheld system developed to digitize a contextual understanding of the scene at a chemical, biological, radiological, nuclear and/or explosives (CBRNE) events is described. The system uses LiDAR and cameras to create a colorized 3D model of the environment, which helps domain experts that are supporting responders in the field. To generate the digitized model, a responder scans any suspicious objects and the surroundings by carrying the system through the scene. The scanning system provides a real-time user interface to inform the user about scanning progress and to indicate any areas that may have been missed either by the LiDAR sensors or the cameras. Currently, the collected data are post-processed on a different device, building a colorized triangular mesh of the encountered scene, with the intention of moving this pipeline to the scanner at a later point. The mesh is sufficiently compressed to be sent over a reduced bandwidth connection to a remote analyst. Furthermore, the system tracks fiducial markers attached to diagnostic equipment that is placed around the suspicious object. The resulting tracking information can be transmitted to remote analysts to further facilitate their supporting efforts. The paper will discuss the system's design, software components, the user interface used for scanning a scene, the necessary procedures for calibration of the sensors, and the processing steps of the resulting data. The discussion will close by evaluating the system's performance on 11 scenes.

Prins, Nicholas↗
DOE OSTI2013

A diffusion synthetic acceleration scheme for rectangular geometries based on bilinear discontinuous finite elements

A DSA technique to accelerate the iterative convergence of S{sub n} transport solves is derived for bilinear discontinuous (BLD) finite elements on rectangular grids. The diffusion synthetic acceleration equations are discretized using BLD elements by adapting the Modified Interior Penalty technique, introduced in [4] for triangular grids. The MIP-DSA equations are SPD and thus are solved using a preconditioned CG technique. Fourier analyses and implementation of the technique in a BLD S{sub n} transport code show that the technique is stable is effective. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

Verification of the pin power reconstruction method in NESTLE

NESTLE is a 3D nodal core simulator that solves the few group neutron diffusion equation using the nodal expansion method (NEM) for the steady state (eigenvalue or fixed-source), adjoint (eigenvalue), and transient (eigenvalue or fixed-source) problems. NESTLE has been in use for over 25 years and has been continuously improved and developed. NESTLE can be used to model either Cartesian geometries (for use with typical PWR, BWR fuel assemblies) or hexagonal geometries (for use with VVER and other triangularly pitched fuel assemblies). As part of NESTLE's ongoing verification and validation process, several computational comparisons have been made to quantify the discrepancies in NESTLE's hexagonal pin power reconstruction method. In this paper, we present the results of these code-to-code comparisons and find that the root mean squared (RMS) error in pin power is comparable to other nodal codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Mu2e - Research & Development

Current efforts are being conducted at Fermi National Laboratory to study potential violations in accepted theory that would otherwise suggest a restructuring of our fundamental understanding of the universe. Mu2e is one of these frontier projects that studies Charged Lepton Flavor Violation (CLFV) which if observed, would suggest physics beyond the Standard Model. Therefore, this note encompasses several projects that contribute to the fruition of Mu2e investigations. Due to the broad range of disciplinary inconsistencies that each project requires, all the work is being presented as a means of justifying contribution to Mu2e. The projects are comprised of a simulation exploring the extinction level of proton pulses after Recycler ring re-bunching by using G4beamline to simulate an 8GeV proton beam interaction with a titanium target, three Cherenkov radiation-based detectors and 2/3-fold and 3/3-fold coincidence rate analysis. Additionally, supplemental work for the implementation of a Micro Telecommunications Computing Architecture (TCA) crate to establish a peak finding algorithm to ensure that the out-of-time beam is less than 10$^{−10}$ fractional level along with single-layer inefficiency analysis on scintillation counters for the Cosmic-Ray Veto (CRV) analysis to ensure the overall inefficiency is 10$^{−4}$. Preliminary results have been achieved for the G4beamline simulation 2/3-fold and 3/3-fold coincidences which are in the order of 10$^{−9}$ and 10$^{−10}$, respectively. Only preliminary results of a triangular counter and four rectangular di-counters for the CRV have been realized. The microTCA crate development is still ongoing.

43 PARTICLE ACCELERATORS↗
DOE OSTI2019

Challenges of SLOWPOKE-2 HEU Irradiated Fuel Transportation from Canada to USA - 19123(2)

The SLOWPOKE-2 (Safe Low-Power Kritical Experiment) was designed by AECL (Atomic Energy of Canada Limited) in the late 1960's. The core consists of HEU (highly enriched uranium) fuel with an enrichment of 93%. The fuel cage typically has approximately 300 fuel pins assembled in a triangular pitch between the two plates. The fuel pins are made from uranium-aluminum alloy cladded with aluminum. The SLOWPOKE-2 reactor is rated at a nominal 20 kW. with varying neutron flux, maximum being 1 E12 n/cm{sup 2}/s. The mass of the fuel cage is 5.06 kg. The life of the core can be extended by adding layers of beryllium reflector. The transportation flask for the irradiated fuel is a Type B(U)F. The shipment classification is UN3328 Class 7. To date, 4 SLOWPOKE-2 reactors have been decommissioned. The fuel from all these reactors was transported to the U.S. The CNSC (Canadian Nuclear Safety Commission) joint Canada-United States guide for approval of Type B(U) and fissile material transportation packages (RD-364) satisfies the requirements stipulated in International Atomic energy Agency (IAEA) TS-R-1 1996 edition (revised) and is acceptable to the DOT (Department of Transportation) and NRC (Nuclear Regulatory Commission) for complying with U.S. code of federal regulations 10 CFR Part 71. The export license is issued by the CNSC. The transportation licence is issued by CNSC after approval from TC (Transport Canada) and also by DOT after approval from NRC. The final home of the HEU SLOWPOKE-2 irradiated fuel is in Savannah River Site of DOE (Department of Energy). The transportation licence requires that the Type B(U)F flask is structurally sound in case of a crash. The US DOT provides a CAC (Competent Authority Certificate) which describes the irradiated fuel specifications that can be transported in the designated flask. The U.S. DOT also requires a leak test certification in accordance with ANSI N14.5-2014 to verify a leakage rate not to exceed 1 E-7 std-cm{sup 3}/s. As most SLOWPOKE-2 facilities are not located on the existing approved routes for nuclear fuel transportation, a route survey needs to be performed and submitted to the NRC. The route survey report is assessed by the NRC and issues a route approval for the transportation of the fissile material. This technical paper will discuss the detailed processes required to obtain approvals in order to transport the HEU SLOWPOKE-2 irradiated fuel from Canada to U.S. based upon recent experiences of decommissioning SLOWPOKE-2 reactors in Canada. It will provide details on the requirements of various regulators in Canada and the U.S. that govern and regulate this shipment including transportation, safety and security. It will also present lessons learned from previous experiences that can be utilized for best practice on other similar projects. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Dynamic Monte Carlo Transient Analysis with Thermal-Hydraulic Feedback by McCARD/CUPID

The reactor core transient analysis requires neutronics calculations with the thermal-hydraulic (TH) feedback. Recently the dynamic Monte Carlo (DMC) neutron transport algorithm controlling delayed neutron precursors was developed and successfully applied for core transient problems with help of a sub-channel code. In the Seoul National University Monte Carlo (MC) code, McCARD, a DMC scheme for the core transient analysis has been implemented by coupling with a computational fluid dynamics code, CUPID using the TCP/IP socket communication. In the McCARD/CUPID transient calculations, temperature-dependent cross sections are produced by a built-in on-the-fly Doppler broadening (OTF DB) module. This paper presents McCARD/CUPID transient analysis results for a triangular unit cell problem in the IAEA coordinated research project (CRP) benchmark on the high temperature gas cooled reactor (HTGR) uncertainty analysis. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Verification of the 2D Finite Element SP3 Solutions

Method(FEM) is the free geometry handling. However many of the geometries of the nuclear engineering problems are not unstructured but rather structured, such as a hexahedral, or triangular prism. Thus, the need for the use of FEM has less portion in nuclear engineering. However it is also certain that it is very hard to exclude the possibility for use of an unstructured geometry such as a PBMR core problem, fusion blanket analysis of the source problem, another design issue with an unstructured geometry. In addition, research regarding the acceleration scheme such as Coarse Mesh Finite Difference (CMFD) showed up in nuclear engineering in recent years. By considering these facts, the FEM cannot be totally exclusive. To utilize the FEM for nuclear engineering problems, several types of assistance are needed. The first is parallelization because of its slow calculation speed. The second is an acceleration scheme. With these two types of assistance the FEM can be utilized freely in the nuclear engineering field. In this research, before conducting parallelization and acceleration, the final verification for the implementation of the up-scattering treatment with an arbitrary geometry conducted through a comparison with the McCARD solution through the well-known C5G7 2D benchmark problem. Note that the earlier work for verification of the FEM for the multi-group problem without up-scattering and with a structured geometry including a mathematical formulation has already been done. The main points that will be tested here are as follows. The first is a check of the right implementation for the multi-group problem with an up-scattering cross section from the perspectives of the multiplication factor and power. The second is the transport effect for this problem by comparing the solutions from P{sub 1} and Simplified P{sub 3}(SP{sub 3}). The third is the effectiveness of the higher order basis function for the same number of unknowns. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

SE2-ANL Subchannel Analysis with New Heating Calculation Based on VARIANT Transport Code

The multi-assembly, steady-state sub-channel analysis code SE2-ANL has been widely used in thermal-hydraulic analysis of sodium cooled fast reactors (SFR). It is used to compute the core-wide temperature profiles for a given flow distribution and to determine the assembly flow rates yielding a desired (assembly-wise) peak temperature distribution. It is a modified version of SUPERENERGY-2 coupled with the DIF3D/GAMSOR/DIF3D procedure for coupled neutron and gamma heating calculations, developed at Argonne National Laboratory (ANL). In the current version of SE2-ANL, the neutron and gamma fluxes are calculated using the finite-difference diffusion theory option of the DIF3D code with six triangular meshes per hexagonal assembly. In addition, the pin power distributions are estimated by assuming that the power distribution within an assembly is separable in the radial and axial directions. Furthermore, the total heat generation rate in an assembly duct wall is simply determined by a user-specified fraction of the total gamma heating within that assembly. In order to determine the pin power distributions more accurately by eliminating these limitations, a new coupled heating calculation procedure has recently developed based on the neutron and gamma transport calculations with the VARIANT code. For this, the GAMSOR code has been extended to generate the intra-nodal gamma source distribution to be consistent with the trial functions of VARIANT. Using the computed neutron and gamma flux distributions, the power distribution in each fuel pin and duct wall is determined using a new utility code CURVE, which has been developed to evaluate the VARIANT flux solution at the center of each fuel pin and along the duct mid-wall. For each pin segment contained in a VARIANT node, the axial power shape is represented as a quadratic profile. The SE2-ANL code has been modified to interface with the new, coupled neutron and gamma heating calculation scheme. In addition, several improvements have been made on the computational methods and models of SE2-ANL. To eliminate the limit on the axial mesh size due to the numerical instability problem of the current explicit scheme, the numerical scheme for discretizing the energy equation in the axial direction has been changed to the {theta}-method of the SLTHEN code. An automated flow allocation scheme has also been implemented in SE2-ANL to determine assembly flow rates such that the peak cladding mid-wall temperatures of individual orifice zones are equalized over the burn cycle. The impacts of the new heating calculation scheme based on VARIANT transport solutions on the cladding and fuel temperatures have been tested against the existing method of SE2-ANL. Using the existing and new heating calculation schemes, full-core subchannel analyses were performed for the metal core design of the 1000 MWt Advanced Burner Reactor (ABR-1000), and the resulting coolant, cladding and fuel temperatures were compared. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Particle Image Velocimetry Measurements in a Wire-Wrapped 61-Pin Hexagonal Fuel Bundle

Liquid metal fast reactor fuel assemblies using sodium as a coolant typically utilize a tightly packed triangular lattice of fuel pins enclosed in a hexagonal duct. The fuel pins are typically wrapped with a helical wire spacer. The primary function of the wire spacers is to maintain a gap between neighboring fuel pins. They also mitigate vortex-induced vibration and enhance sub-channel mixing to increase convective heat transfer. Many experiments have been performed to study thermal-hydraulic characteristics, primarily pressure drops, of the wire-wrapped fuel bundles. Novendstern developed a semi-empirical model to predict pressure losses in a hexagonal array of wire-wrapped pins in the turbulent flow regime. Rehme proposed pressure drop correlations for a wide range of geometrical parameters, such as pitch-to-diameter ratio P/D from 1.125 to 1.417, lead-to-wire diameter ratio H / dm from 6 to 45, and number of pins n from 7 to 61. Cheng and Todreas introduced their hydrodynamic models for subchannel friction factors and mixing parameters in a 37-pin fuel bundle, which covered the laminar, transition and turbulent flow regimes. Computational fluid dynamics (CFD) offers an opportunity to validate the above correlations and investigate the flow characteristics in the subchannels. Several numerical studies of the wire-wrapped fuel bundles have been performed and most of them are Reynolds-Averaged Navier-Stokes (RANS) based simulations with various turbulence models. Ahmad and Kim performed RANS simulations of 7- and 19-pin bundles using the k-omega SST turbulent model, while Gajapathy et al. performed RANS simulations on 7-pin bundles using the k-epsilons model. Recently, Fischer et al. carried out large-eddy simulation (LES) calculations on wire-wrapped fuel bundles to obtain higher-resolution numerical solutions of subchannel behavior. The geometrical complexity of the hexagonal wire-wrapped fuel bundle has limited experimental activities to pressure drop measurements. A database of pressure drops and flow-field measurements in a 61-pin wire-wrapped hexagonal fuel bundle is not available to benchmark the existing correlations and validate the CFD calculations. Texas A and M University is conducting isothermal flow experiments in a wire-wrapped 61-pin hexagonal fuel bundle to support the research on advanced nuclear fuel development sponsored by the Department of Energy (DOE). The purpose of these tests is to perform high spatial and temporal resolution measurements of the flow and pressure measurements at different locations in the fuel bundle. The experimental facility employs matched-index-of-refraction (MIR) techniques and laser diagnostic velocity measurement techniques such as particle image velocimetry (PIV) and laser Doppler velocimetry (LDV). The result of this experimental activity will be an experimental database of pressure and flow-field measurements that will be suitable for validating CFD computer codes. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Spacer Grid and Mixing Vanes Effects Study on Two-phase Flow Using Bubble Tracking Simulations

In Light Water Reactor (LWR) cores, the nuclear fuel rods are arranged together as rod bundles in a triangular or square pattern. The fuel rod bundles are supported by the spacer grids with mixing vanes. An understanding of spacer grid and mixing vanes effect is important since it has a direct impact on the turbulence structure in the flow and the heat removal efficiency from the fuel rod surface. Although a considerable amount of literature has been published on the experimental investigation of spacer grid and mixing vanes effect, the measurements were mainly focused on single-phase flow. The influence of spacer grid and mixing vanes on two-phase flow is yet to be fully understood. Some preliminary research efforts have been invested in nuclear community. For example, Wheeler et al. experimentally studied the effects of spacer grid in air-water two-phase flow. Although mixing vanes are not considered, they presented interesting results of time-averaged two-phase flow quantities at upstream and downstream locations of the spacer grid. Over the past decades, the rapid development of High-Performance Computing (HPC) is allowing for the increasingly larger-scale high-fidelity simulations. Thanks to this favorable historical trend, Direct Numerical Simulation (DNS), integrated with Interface Tracking Methods (ITM), has been emerging as a valuable tool to complement and expand our ability to understand the two-phase flow phenomenon. Meanwhile, in order to make best use of the large-scale two-phase flow simulations, advanced data analysis techniques are being developed, as they enable the extraction of detailed parameters about individual bubble behavior. The level set ITM is utilized in our DNS flow solver, which makes use of a signed distance field. The gas-liquid interface is modeled by the zero level set. The level set method can be readily used to distinguish different phases based on the sign of corresponding level set value (e.g. the sign is positive in the liquid phase while negative in the gas phase.). However, the traditional level set approach is not able to collect calculated values and associate with specific bubbles when multiple (even thousands of) bubbles present in the simulations of two-phase flow. This drawback hinders the collection of useful bubble information, which can give us valuable insights about bubbly flows. For example, how the different local fluid conditions could affect bubble interfacial forces, bubble deformation level, and eventually the bubble distribution throughout the whole domain. A bubble tracking methodology has been developed to address the drawbacks and improve the data collection efficiency of level set based two-phase flow simulations. In the present work, a single Pressurized Water Reactor (PWR) subchannel with reduced size spacer grid and mixing vanes is chosen as the computational domain. The springs and dimples of the spacer grid are not represented for simplicity in the present investigation. The bubble tracking methodology is applied to the simulations of two-phase bubbly flow through spacer grid and mixing vanes region. In the intended simulation, the liquid turbulence will be fully resolved by DNS while the two-phase behaviors are captured by level set ITM. Bubble tracking data are to be processed to investigate the spacer grid and mixing vanes effect on two-phase bubbly flow. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗