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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 1,405 records · Page 78
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Phase stability and microstructure of neutron-irradiated substoichiometric yttrium dihydrides

The impact of the neutron-displacement damage on phase stability and microstructure of substoichiometric yttrium dihydrides (YH x , x <2) were investigated to assess their use as solid moderator in high-temperature nuclear reactors. YH x specimens were, thus, subjected to neutron irradiations in the range of 0.1–2 displacements per yttrium atom (dpa-Y) in the temperature range of 536–878°C at the Oak Ridge National Laboratory's (ORNL's) High Flux Isotope Reactor (HFIR). YH x specimens were initially prepared at stoichiometry (H/Y) ratios of 1.69 and 1.83. HFIR-irradiated specimens were characterized by variety of techniques to investigate H retention characteristics including dimensional analysis, optical microscopy, scanning electron microscopy electron back scatter diffraction (EBSD), transmission electron microscopy, thermal desorption spectroscopy (TDS), and high-energy x-ray diffraction (HE-XRD) characterizations. Overall, YH x exhibited notable structural and phase stability under short-term neutron-irradiation, except for the samples with significant silicon carbide (SiC) interaction at high doses and temperatures. Basic dimensional and mass measurements were misleading for accurate assessment of H retention, as confirmed by EBSD phase maps, XRD line profiles, and TDS signals. Thus, it was discussed that a robust H retention metric is needed to assess irradiated hydrides. Further, nanoscale cavities were observed as a result of the neutron irradiation in all samples. Although no clear impact of dose and irradiation temperature was determined, the initial H/Y ratio had an impact on the cavity number density where low H/Y specimens had high-resistance to cavity formation. The Y-vacancy cluster formation at the collision stage of the displacement cascade and their stabilization by H were considered to be the likely underlying mechanisms for the observed cavity microstructure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Machine tool cross beam design, fabrication, and testing using metal big area additive manufacturing

This paper describes the application of metal Big Area Additive Manufacturing (mBAAM) to the fabrication of a machine tool cross beam. The replacement of a traditional box design weldment with a new design printed by wire arc additive manufacturing using the MedUSA system at Oak Ridge National Laboratory (ORNL) is detailed. This requires a new design strategy based on the unique mBAAM capabilities. The intent of the new design is to reduce mass, while maintaining the dynamic stiffness. To compare the two designs, the natural frequencies and mode shapes are measured using impact testing and predicted using finite element analysis. It is confirmed that the printed structure dynamics agreed with the numerical model predictions, which demonstrates that it is feasible to model a large-scale mBAAM part and understand its behavior prior to printing. Another notable outcome of this study is that the significant residual stress and distortion in the print indicate that knowledge gaps remain for widespread implementation of mBAAM.

42 ENGINEERING↗
DOE OSTI2020

Dynamic phase transformations in additively manufactured Ti-6Al-4V during thermo-mechanical gyrations

A complex interaction of process parameters, geometry and scan strategies in Additive Manufacturing (AM), can bring about spatial and temporal transients, i.e., Σ T ( x, y, z, time ), within a part. Published literature focusses on fluctuating thermal cycles on the microstructure evolution. However, the microstructural variations have not been correlated to dynamic flow behavior due to the macro- and micro-scale phenomena, i.e., accumulated plastic strains brought about by large thermal gradients, transformational strains and crystallographic misfit strains. Therefore, here we studied the mechanical response of Ti6Al4V alloys produced by AM under externally imposed controlled thermo-mechanical reversals in a Gleeble® thermo-mechanical simulator. The stress-strain behaviors were correlated to phase fractions, lattice strains, and also limited information on crystallographic texture using neutron diffraction techniques at the VULCAN Beamline at SNS, ORNL and also metallographic studies. The results are discussed and rationalized based on theories of static and dynamic phase transformations.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Nuclear Data Sheets for A = 87

The evaluated experimental data are presented for 14 known nuclides of mass 87 (Ga, Ge, As, Se, Br, Kr, Rb, Sr, Y, Zr, Nb, Mo, Tc, Ru). New datasets have been added for {sup 87}Ga, {sup 87}Se, {sup 87}Kr, {sup 87}Rb, {sup 87}Sr, {sup 87}Nb, and {sup 87}Tc. The observation of and new lifetime measurements for {sup 87}As at the NSCL Coupled Cyclotron Facility and the Oak Ridge Isochronous Cyclotron at the HRIBF–ORNL Facility are included in the adopted levels for this isotope. The new evaluation also includes an isomeric level at unknown energy for {sup 87}Tc, changes to the adopted levels for {sup 87}Nb based on newly published data including an extension of the high spin levels and new lifetime measurements, and new heavy ion data for {sup 87}Kr. New high spin levels for {sup 87}Rb were observed, indicating possible neutron core excitations. In addition, the ground state lifetime of {sup 87}Rb has been revised and a new lifetime measurement for an excited level in {sup 87}Rb indicates a possible proton spin flip. New excitation levels have been observed for the first time in {sup 87}Se. New direct and precise measurement of atomic masses of {sup 87}Tc, {sup 87}Mo, {sup 87}Nb, {sup 87}Zr, {sup 87}Mo and {sup 87}Tc have greatly improved the Q value landscape in this mass region. This work supersedes the previous A=87 evaluation published in 2002He09, and Q values for all isotopes were updated to incorporate the mass evaluation in 2012Wa38.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

The nonlinear light output of NaI(Tl) detectors in the Modular Total Absorption Spectrometer

New detector array, the Modular Total Absorption Spectrometer (MTAS),was commissioned at the Holifield Radioactive Ion Beam Facility (HRIBF) at Oak Ridge National Lab(ORNL).Total absorption gamma spectra measured with MTAS are expected to improve beta-feeding patterns and beta strength functions in fission products.MTAS is constructed out of hexagonal NaI(Tl) detectors with a unique central module surrounded by 18 identical crystals assembled in three rings. The total NaI(Tl) mass of MTAS is over1000 kg.The response of the central and other 18 MTAS modules to -radiation was simulated using the GEANT4 tool kit modified to analyze the nonlinear light output of NaI(Tl).A detailed description oftheGEANT4modifications madeisdiscussed.SimulatedenergyresolutionofMTAS modules is found to agree well with the measurements for single transitions of 662keV (137Cs) with 8.2% full width half maximum (FWHM),835keV (54Mn) with FWHM of 7.5% FWHM, and 1115keV (65Zn) with FWHM of 6.5%.Simulations of single and multiple -rays from 60Co are also discussed.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2012

FABRICATION OF URANIUM OXYCARBIDE KERNELS AND COMPACTS FOR HTR FUEL

As part of the program to demonstrate tristructural isotropic (TRISO)-coated fuel for the Next Generation Nuclear Plant (NGNP), Advanced Gas Reactor (AGR) fuel is being irradiation tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). This testing has led to improved kernel fabrication techniques, the formation of TRISO fuel particles, and upgrades to the overcoating, compaction, and heat treatment processes. Combined, these improvements provide a fuel manufacturing process that meets the stringent requirements associated with testing in the AGR experimentation program. Researchers at Idaho National Laboratory (INL) are working in conjunction with a team from Babcock and Wilcox (B&W) and Oak Ridge National Laboratory (ORNL) to (a) improve the quality of uranium oxycarbide (UCO) fuel kernels, (b) deposit TRISO layers to produce a fuel that meets or exceeds the standard developed by German researches in the 1980s, and (c) develop a process to overcoat TRISO particles with the same matrix material, but applies it with water using equipment previously and successfully employed in the pharmaceutical industry. A primary goal of this work is to simplify the process, making it more robust and repeatable while relying less on operator technique than prior overcoating efforts. A secondary goal is to improve first-pass yields to greater than 95% through the use of established technology and equipment. In the first test, called “AGR-1,” graphite compacts containing approximately 300,000 coated particles were irradiated from December 2006 to November 2009. The AGR-1 fuel was designed to closely replicate many of the properties of German TRISO-coated particles, thought to be important for good fuel performance. No release of gaseous fission product, indicative of particle coating failure, was detected in the nearly 3-year irradiation to a peak burn up of 19.6% at a time-average temperature of 1038–1121°C. Before fabricating AGR-2 fuel, each fabrication process was improved and changed. Changes to the kernel fabrication process included replacing the carbon black powder feed with a surface-modified carbon slurry and shortening the sintering schedule. AGR-2 TRISO particles were produced in a 6-inch diameter coater using a charge size about 21-times that of the 2-inch diameter coater used to coat AGR-1 particles. The compacting process was changed to increase matrix density and throughput by increasing the temperature and pressure of pressing and using a different type of press. AGR-2 fuel began irradiation in the ATR in late spring 2010.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Detection and analysis of particles with failed SiC in AGR-1 fuel compacts

As the primary barrier to release of radioactive isotopes emitted from the fuel kernel, retention performance of the SiC layer in tristructural isotropic (TRISO) coated particles is critical to the overall safety of reactors that utilize this fuel design. Most isotopes are well-retained by intact SiC coatings, so pathways through this layer due to cracking, structural defects, or chemical attack can significantly contribute to radioisotope release. In the US TRISO fuel development effort, release of 134 Cs and 137 Cs are used to detect SiC failure during fuel compact irradiation and safety testing because the amount of cesium released by a compact containing one particle with failed SiC is typically ten or more times higher than that released by compacts without failed SiC. Compacts with particles that released cesium during irradiation testing or post-irradiation safety testing at 1600–1800 °C were identified, and individual particles with abnormally low cesium retention were sorted out with the Oak Ridge National Laboratory (ORNL) Irradiated Microsphere Gamma Analyzer (IMGA). X-ray tomography was used for three-dimensional imaging of the internal coating structure to locate low-density pathways through the SiC layer and guide subsequent materialography by optical and scanning electron microscopy. In addition, all three cesium-releasing particles recovered from as-irradiated compacts showed a region where the inner pyrocarbon (IPyC) had cracked due to radiation-induced dimensional changes in the shrinking buffer and the exposed SiC had experienced concentrated attack by palladium; SiC failures observed in particles subjected to safety testing were related to either fabrication defects or showed extensive Pd corrosion through the SiC where it had been exposed by similar IPyC cracking.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Thermal hydraulics modeling of On-line refueling for the Advanced High Temperature Reactor (AHTR)

As part of the effort to move forward the Fluoride salt-cooled High-temperature Reactor (FHR) technology, the Advanced High Temperature Reactor (AHTR) is being developed at Oak Ridge National Laboratory (ORNL) and several other institutions. Due to its plank fuel design and inherently low heavy metal loading, which challenges fuel utilization, online refueling is considered as an option for improving the economic viability of this reactor. This paper presents the thermal hydraulic modeling of the online refueling for the AHTR. Analyses at the single channel and fuel assembly level have been performed in order to develop a model of the reactor. The reactor modeling has been integrated with CFD studies for characterizing the steady state conditions and the refueling transient. The simulations of the transient have been complemented by the analysis of the stability of the refueling operation. This work demonstrates the viability of online refueling from a thermal hydraulic standpoint and develops a modeling approach for this type of operational transient.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Development and Implementation of a CTF Code Verification Suite.

CTF is a thermal hydraulic subchannel code developed to predict light water reactor (LWR) core behavior. It is a version of Coolant Boiling in Rod Arrays (COBRA) developed by Oak Ridge National Laboratory (ORNL) and North Carolina State University (NCSU) and used in the Consortium for the Advanced Simulation of LWRs (CASL). Here, the existing CTF code verification matrix is expanded, which ensures that the code is a faithful representation of the underlying mathematical model. The suite of code verification tests are mapped to the underlying conservation equations of CTF and significant gaps are addressed. As such, five new problems are incorporated: isokinetic advection, conduction, pressure drop, convection, and pipe boiling. Convergence behavior and numerical errors are quantified for each of the tests and all tests converge at the correct rate to their corresponding analytic solution. A new verification utility that generalizes the code verification process is used to incorporate these problems into the CTF automated test suite.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

CFD analysis of a solid pin-fueled small modular fluoride salt-cooled reactor

A computational fluid dynamics (CFD) study is performed for the pre-conceptual solid pin-fueled small modular fluoride salt-cooled reactor developed by the Oak Ridge National Laboratory (ORNL). The solid and fluid regions in a 1/12th section of a fuel assembly are modeled. The solid region includes fuel and non-fuel pins, with heat generation in the active region of the fuel pins. Two different power profiles, uniform and center-peaked, are considered. The FLiBe coolant flows from the bottom to the top of the core, parallel to the bank of fuel and non-fuel pins. The k-ω shear-stress transport (SST) model is chosen as the baseline turbulence model. The effects of grid refinement, inlet turbulence specification, and turbulence models on the temperature and pressure drop predictions are studied. Turbulence model sensitivity is investigated by comparing the results from the k-ω SST model with other two-equation models (k-ω baseline or BSL and k-ε realizable) as well as with anisotropic Reynolds stress transport models (linear pressure-strain and stress-BSL). Here the results show that the choice of turbulence model has a significant impact on the pin temperature and bundle pressure drop predictions. The results from the baseline turbulence model show good agreement between the fuel pin temperatures and bundle pressure drop values predicted using a subchannel model previously developed by the authors of the present study.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

A whole-core steady-state thermal-hydraulic model for annular fuel type fluoride-salt-cooled reactors

A whole-core, steady-state thermal-hydraulic model is developed for the fluoride-salt-cooled small modular advanced high-temperature reactor (SmAHTR) that employs an annular fuel configuration. This pre-conceptual reactor design by Oak Ridge National Laboratory (ORNL) has the annular fuel and moderator pins arranged in a hexagonal layout. The FLiBe coolant flows from the bottom to the top of the core, parallel to the hexagonal bundle. The fuel and moderator pins in the core are discretized into finite volumes and the 3-D heat conduction equation is solved to obtain the temperature profile. Inter-fuel assembly conduction is also addressed. For this fuel assembly configuration, the coolant flows through two distinct regions – the hexagonal pin bundle and the annulus between the fuel pin and the tie rod. The fluid flow through the hexagonal bundles is modeled using the subchannel approach, in which the coolant region is discretized into corner, edge and interior subchannels and the resulting conservation equations are systematically solved. The 1-D mass, momentum and energy equations are solved for the annulus channels between the fuel pin and the tie rod. Pertinent closure models from the literature are employed to close the system of equations. We also performed a preliminary code-to-code comparison between the present model and a CFD model.. The resulting thermal-hydraulic model can provide temperature, flow rate and pressure drop profiles for the different solid and fluid regions throughout the entire core. Whole-core thermal-hydraulic results for a representative power profile are presented and discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2026

Numerical implementation of the MOOSE subchannel module (SCM) algorithm

The subchannel module of moose (SCM), previously referred to as Pronghorn-SC, is a subchannel code designed to resolve single-phase flow fields and calculate the relevant flow variables, in nuclear reactor fuel-pin assemblies. The assemblies it models are: water-cooled, with bare fuel-pins in a square ducted, quadrilateral lattice and liquid-metal/water-cooled, with bare/wire-wrapped fuel pins in a hexagonal ducted, triangular lattice. Previous publications have presented the development, validation and verification of SCM. This work presents an comparative overview of the different solvers implemented within SCM. This includes the solver algorithm description, and a performance comparison. The two test cases chosen to demonstrate the solver performance were taken from the PSBT enthalpy mixing benchmark and the ORNL-19 pin benchmark.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Parallel accelerated vector similarity calculations for genomics applications

The surge in availability of genomic data holds promise for enabling determination of genetic causes of observed individual traits, with applications to problems such as discovery of the genetic roots of phenotypes, be they molecular phenotypes such as gene expression or metabolite concentrations, or complex phenotypes such as diseases. However, the growing sizes of these datasets and the quadratic, cubic or higher scaling characteristics of the relevant algorithms pose a serious computational challenge necessitating use of leadership scale computing. In this study we describe a new approach to performing vector similarity metrics calculations, suitable for parallel systems equipped with graphics processing units (GPUs) or Intel Xeon Phi processors. Our primary focus is the Proportional Similarity metric applied to Genome Wide Association Studies (GWAS) and Phenome Wide Association Studies (PheWAS). We describe the implementation of the algorithms on accelerated processors, methods used for eliminating redundant calculations due to symmetries, and techniques for efficient mapping of the calculations to many-node parallel systems. Finally, results are presented demonstrating high per-node performance and parallel scalability with rates of more than five quadrillion (5 × 10 15 ) elementwise comparisons achieved per second on the ORNL Titan system. In a companion paper we describe corresponding techniques applied to calculations of the Custom Correlation Coefficient for comparative genomics applications.

59 BASIC BIOLOGICAL SCIENCES↗
DOE OSTI2022

Low-synch Gram–Schmidt with delayed reorthogonalization for Krylov solvers

The parallel strong-scaling of iterative methods is often determined by the number of global reductions at each iteration. Low-synch Gram-Schmidt algorithms are applied here to the Arnoldi algorithm to reduce the number of global reductions and therefore to improve the parallel strong-scaling of iterative solvers for nonsymmetric matrices such as the GMRES and the Krylov-Schur iterative methods. In the Arnoldi context, the factorization is "left-looking" and processes one column at a time. Among the methods for generating an orthogonal basis for the Arnoldi algorithm, the classical Gram-Schmidt algorithm, with reorthogonalization (CGS2) requires three global reductions per iteration. A new variant of CGS2 that requires only one reduction per iteration is presented and applied to the Arnoldi algorithm. Delayed CGS2 (DCGS2) employs the minimum number of global reductions per iteration (one) for a one-column at-a-time algorithm. The main idea behind the new algorithm is to group global reductions by rearranging the order of operations. DCGS2 must be carefully integrated into an Arnoldi expansion or a GMRES solver. Numerical stability experiments assess robustness for Krylov-Schur eigenvalue computations. Performance experiments on the ORNL Summit supercomputer then establish the superiority of DCGS2 over CGS2.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

New Polarized Small Angle Neutron Scattering capability at the High Flux Isotope Reactor

Polarized Small Angle Neutron Scattering (SANS) is a powerful tool for studying mesoscale magnetic structures. In a recent development at the Oak Ridge National Laboratory (ORNL), a polarized neutron beam has been implemented for the CG-2 (GP-SANS) beamline at the High Flux Isotope Reactor (HFIR). Three major components were developed to provide the polarized neutron capability: a transmission supermirror neutron polarizer, an adiabatic fast passage neutron spin flipper and a polarized 3 He filter as neutron spin analyzer. Finally, the new polarized beam SANS capability was used to investigate several magnetic samples having mesoscale structures.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2013

ParCAT: Parallel Climate Analysis Toolkit

Climate science is employing increasingly complex models and simulations to analyze the past and predict the future of Earth's climate. This growth in complexity is creating a widening gap between the data being produced and the ability to analyze the datasets. Parallel computing tools are necessary to analyze, compare, and interpret the simulation data. The Parallel Climate Analysis Toolkit (ParCAT) provides basic tools to efficiently use parallel computing techniques to make analysis of these datasets manageable. The toolkit provides the ability to compute spatio-temporal means, differences between runs or differences between averages of runs, and histograms of the values in a data set. ParCAT is implemented as a command-line utility written in C. This allows for easy integration in other tools and allows for use in scripts. This also makes it possible to run ParCAT on many platforms – from laptops to supercomputers. ParCAT outputs NetCDF files so it is compatible with existing utilities such as Panoply and UV-CDAT. This paper describes ParCAT and presents results from some example runs on the Titan system at ORNL.

97 MATHEMATICS AND COMPUTING↗