Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Research reactor”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 523 records · Page 29
All sourcesNASA NTRSDOE OSTI
NASA NTRS1984

Silicon refinement by chemical vapor transport

Silicon refinement by chemical vapor transport is discussed. The operating characteristics of the purification process, including factors affecting the rate, purification efficiency and photovoltaic quality of the refined silicon were studied. The casting of large alloy plates was accomplished. A larger research scale reactor is characterized, and it is shown that a refined silicon product yields solar cells with near state of the art conversion efficiencies.

Olson, J.↗
NASA NTRS1991

Laser velocimetry measurements in non-isothermal CVD systems

Researchers at the NASA Langley Research Center are applying laser velocimetry (LV) techniques to characterize the fluid dynamics of non-isothermal flows inside fused silica chambers designed for chemical vapor deposition (CVD). Experimental issues involved in the application of LV techniques to this task include thermophoretic effects on the LV seed particles, seeding the hazardous gases, index of refraction gradients in the flow field and surrounding media, optical access, relatively low flow velocities, and analysis and presentation of sparse data. An overview of the practical difficulties these issues represent to the use of laser velocimetry instrumentation for CVD applications is given. A fundamental limitation on the application of LV techniques in non-isothermal systems is addressed which involves a measurement bias due to the presence of thermal gradients. This bias results from thermophoretic effects which cause seed particle trajectories to deviate from gas streamlines. Data from a research CVD reactor are presented which indicate that current models for the interaction of forces such as Stokes drag, inertia, gravity, and thermophoresis are not adequate to predict thermophoretic effects on particle-based velocimetry measurements in arbitrary flow configurations.

Johnson, E. J.↗
NASA NTRS1991

Flow field velocity measurements for non-isothermal systems

Experimental techniques which can be potentially utilized to measure the gas velocity fields in nonisothermal CVD systems both in ground-based and space-based investigations are considered. The advantages and disadvantages of a three-component laser velocimetry (LV) system that was adapted specifically for quantitative determination of the mixed convective flows in a chamber for crystal growth and film formation by CVD are discussed. Data from a horizontal research CVD reactor indicate that current models for the effects of thermophoretic force are not adequate to predict the thermophoretic bias in arbitrary flow configurations. It is concluded that LV techniques are capable of characterizing the fluid dynamics of a CVD reactor at typical growth temperatures. Thermal effects are shown to dominate and stabilize the fluid dynamics of the reactor. Heating of the susceptor increases the gas velocities parallel to the face of a slanted susceptor by up to a factor of five.

Johnson, E. J.↗
NASA NTRS2023

Galvanic Protected Anode Materials for Molten Regolith Electrolysis

This effort aimed to explore the use of refractory alloys as long-lifetime electrode materials for use in a molten regolith electrolysis (MRE) reactor, a technology that can produce oxygen (O2) from regolith. To find a material capable of withstanding the high temperatures and corrosive nature of molten regolith, first-principles thermodynamic calculations were performed on platinum group metals in the presence of iron (Fe) and silicon (Si) to determine favorable compositions for resisting unwanted secondary phases that could form on the surface of the material. As a result of these first-principles calculations, the kinetic properties of elements such as aluminum (Al), Fe, and Si can be more easily derived, which would provide insight into the longevity of such a material when used in an MRE reactor. The effort funded by this CIF project mapped the entire phase diagram of platinum group metals, which is beneficial not only to MRE where high temperatures and molten material are present, but also aerospace applications like turbine materials, terrestrial metal processing techniques, and energy storage applications. This project narrowed a broad spectrum of refractory metals down to a small list of alloys that would be able to survive not only the high temperatures of an MRE reactor, but also the thermodynamic forces that would cause inferior alloys to form secondary phases, which would ultimately consume the electrodes of the reactor. This research needs to be followed up with experimental demonstrations of these alloys in an MRE environment.

Kevin Grossman↗
DOE OSTI2021

Material transfers detection with seismic observations

We are exploring the use of data from a seismic network around a research nuclear reactor and isotope production facility at Oak Ridge National Laboratory to study activity patterns related to the transfer of nuclear material. A sensor network with eight seismometers was installed around the High Flux Isotope Reactor and the Radiochemical Engineering Development Center and started operating in October 2019. These data are used to detect and characterize operational events around facility. We are using those data to extract signals related to the movement of vehicles involved in the transport of nuclear materials (e.g., transportation of reactor fuel, targets, and product isotopes). Large vehicles produce mechanical energy that can be observed as seismic signals, either as the result of sound emanating from the vehicle engine or by the generation of surface waves as a response of the ground to the load of the vehicles. In particular, the station located near the entrance gate of the facility displays a clear seismic signature as vehicles cross a metal platform on the ground. This signal is characterized by sharp energy bursts that correspond to the number of axles on the vehicle. We developed and are testing an algorithm to identify sequences of energy bursts to count and characterize vehicles of different sizes. We are also exploring the use of seismic polarization analysis to measure the degree of polarization of the observed signals. These two techniques for detection and characterization of vehicles used for nuclear material transfers will be validated with seismic data from a targeted collection using different vehicles and routes similar to the ones used in real-life scenarios at the High Flux Isotope Reactor.

Marcillo, Omar↗
DOE OSTI2022

UPDATES FROM THE INVOLUTE WORKING GROUP

The HFIR, RHF, and FRM II reactors represent a particular class of Research and Test Reactors that provide some of the most intense and continuous neutron fluxes for science, industry, and medical applications. These high-performance reactors have achieved compact cores by operating with Highly Enriched Uranium fuel (HEU, 235U/U ≥ 20 wt. %) and utilizing fuel plates curved as an involute. Due to the proliferation risks, the international community aims to reduce or eliminate, when possible, the use of HEU fuel in civilian facilities by converting them to a Low-Enriched Uranium fuel (LEU, 235U/U < 20 wt. %). Conversion of these reactors without significantly compromising their performance or safety is a challenging endeavor that can tremendously benefit from advanced computational tools and thus, eliminate unnecessary conservatism to ensure sufficient thermal margins. Therefore, models are being developed using modern Computational Fluid Dynamics (CFD) and Computational Structural Mechanics (CSM) software to evaluate the steady-state safety margins of various LEU designs instead of being reliant on the more traditional, conservative methods. To gain the confidence and acceptance of high-fidelity modeling by the nuclear regulators, Argonne National Laboratory (ANL) and the involute reactors have formed an informal scientific group, the Involute Working Group (IWG). The IWG facilitates inter-organizational collaboration on experimental benchmarking, code-to-code comparisons, and Verification and Validation (V&V). This paper describes some of the recent IWG efforts in validating software against the existing experimental data, as well as code-to-code comparisons of different software used by the IWG members.

Bergeron, Aurelien↗
DOE OSTI2022

Uncertainty Quantification of Calculated Temperatures for the AGR 5/6/7 Experiment

This report documents the quantification of uncertainty of the calculated temperature data for the Advanced Gas Reactor (AGR) 5/6/7 fuel irradiation experiment conducted in the Advanced Test Reactor at Idaho National Laboratory in support of the Advanced Reactor Technologies? research and development program. Recognizing uncertainties inherent in physics and thermal simulations of the AGR 5/6/7 capsules, the results of the numerical simulations are used in combination with statistical analysis methods to improve qualification of measured data. The calculated fuel temperatures for AGR tests are also used for validation of the fission product transport and fuel performance simulation models. These crucial roles of the calculated fuel temperatures in ensuring achievement of the AGR experimental program objectives require accurate determination of the model temperature uncertainties. This report covers temperature uncertainty results for each of the five AGR 5/6/7 capsules. To quantify the uncertainty of calculated temperatures determined using the ABAQUS finite element heat transfer code, this study identifies and analyzes model parameters of potential importance to the calculated temperatures of fuel compacts and thermocouples. The selection of input parameters for uncertainty quantification is based on the ranking of their influences upon temperature predictions. Thus, selected input parameters include those with high sensitivity and those with the largest uncertainty. Propagation of model parameter uncertainty and sensitivity is then used to quantify the overall uncertainty of calculated temperatures. Measurement uncertainty, analysis of modeling assumptions, and expert judgment are used as the basis to quantify the uncertainty range for selected input parameters. The input uncertainties are dynamic, accounting for the effect of unplanned events and changes in thermal properties of capsule components over extended exposure to high temperatures and fast neutron irradiation. The sensitivity analysis performed in this work went beyond the traditional local sensitivity. Using experimental design, analysis of pairwise interactions of model parameters was performed to establish sufficiency of the time dependent first order (linear) expansion terms in constructing the temperature response surface. To achieve completeness, uncertainty propagation made use of pairwise noise correlations of model parameters. Furthermore, using an interpolation scheme over the input parameter domain, the analysis obtains time dependent sensitivity over the test campaign duration. This allows computation of uncertainty for the calculated fuel temperatures and the calculated graphite temperatures at thermocouple locations during the entire irradiation period.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

October 2022 NS&T Highlights

These are the highlights from Nuclear Science and Technology for the month of October.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

February 2023 NS&T Highlights

Monthly highlights for February for the Nuclear Science & Technology directorate

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

October 2023 NS&T Highlights

An achievement report for the month of October 2023 for the Nuclear Science & Technology directorate.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Evaluation of the NSUF Reactor Activation and Damage (RAD) Calculator damage component

The Nuclear Science User Facilities (NSUF) provide researchers access to nuclear research facilities and expertise all over the United States and at the Belgian Center for Nuclear Research (SCK/CEN). One special capability of the NSUF is providing access to nuclear research and test reactors often used for conducting material testing experiments. These services are provided at no cost to the researcher through a competitive, peer-reviewed process. With so many proposals being submitted simultaneously, an NSUF information management resource called the Combined Materials Experiment Toolkit (CoMET) was implemented to speed up the review process and give users access to the information and tools necessary for designing meaningful experiments and writing the best proposals possible. One of CoMET’s tools is the Reactor Activation and Damage (RAD) Calculator. For those with little or no background in it, neutron irradiation in research and test reactors can be a challenging subject to include in a proposal. That is why the RAD Calculator was created to assist users with providing estimates of irradiation and post-irradiation conditions, enabling them to select the best neutron irradiation and post-irradiation examination facilities for reaching their experiment goals. The RAD Calculator is a general scoping tool that estimates radiation damage in terms of DPA, in addition to the resulting radioactivity of the material post-irradiation. This report evaluates the neutron damage component of the RAD Calculator. The evaluation of the activation component of the calculator is covered in report INL/EXT-20-58080.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1987

Two-phase reduced gravity experiments for a space reactor design

Future space missions researchers envision using large nuclear reactors with either a single or a two-phase alkali-metal working fluid. The design and analysis of such reactors require state-of-the-art computer codes that can properly treat alkali-metal flow and heat transfer in a reduced-gravity environment. New flow regime maps, models, and correlations are required if the codes are to be successfully applied to reduced-gravity flow and heat transfer. General plans are put forth for the reduced-gravity experiments which will have to be performed, at NASA facilities, with benign fluids. Data from the reduced-gravity experiments with innocuous fluids are to be combined with normal gravity data from two-phase alkali-metal experiments. Because these reduced-gravity experiments will be very basic, and will employ small test loops of simple geometry, a large measure of commonality exists between them and experiments planned by other organizations. It is recommended that a committee be formed to coordinate all ongoing and planned reduced gravity flow experiments.

Antoniak, Zenen I.↗
DOE OSTI2024

Uncertainty Quantification and Sensitivity Analysis of Non-Nuclear Advanced Controls Testbed Reactor Mockup

The research presented in this report describes our progress in applying stochastic methods and uncertainty quantification, parametric study, and variance-based sensitivity analysis (also known as Sobol sensitivity analysis) to a full-core model of a nuclear thermal propulsion (NTP) system simulated with Griffin, with the goal of developing a reduced order (surrogate) model which can be rapidly sampled while perturbing multiple input parameters. In this NTP system, reactivity and power feedback affect the rotation of control drums, which are controlled by a hybrid proportional, integral and derivative (PID) controller, actuated by the power demand and reactivity feedback from the numerical model. This model uses reactor kinetic feedback (mean generation time and $\beta$ from a transient Griffin simulation executed with the improved quasi-static method to provide the kinetic parameters) as inputs to functions which control the CD rotation angle. Using a number of stochastic method approaches, we developed a dual purpose training-surrogate model of the NTP system using polynomial regression. The trained model can be rapidly sampled while simultaneously perturbing various input parameters of the model, such as coefficients on the PID control, or temperature (directly affect the neutron cross section). The surrogate model delivers accurate results orders-of-magnitude faster (minutes, not days) than the base model. Once the base model has been trained, distributions of the uncertain parameters can be changed at will to investigate the effects of perturbing multiple inputs and their effect on the output. For example, coefficients used in the PID control system may vary due to some physical interference, or there may be uncertainty in the temperature of the neutron cross sections in various regions of the reactor. A distribution can be placed on these parameters and operational boundaries can be determined. The goal of this work is to support development of an advanced control system to operate CDs in a functioning NTP system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Micro-tensile testing of the bond line in hot isostatic pressed aluminum

Considerable effort is being devoted to development and regulatory qualification of low enriched fuels for research and test reactors by many agencies worldwide. One promising fuel configuration being examined for United States higher power research and test reactors (USHPRRs) are plate-type fuels composed of a metallic uranium-molybdenum foil clad in an aluminum alloy. The two pieces of aluminum alloy cladding are bonded using a hot isostatic pressing method. The mechanical properties of the resulting bond line in the aluminum alloy cladding will vary by the HIP'ing parameters, requiring a need to characterize the bond line. Small scale mechanical testing can provide a path for evaluating the mechanical properties and deformation behavior of the bond line both prior to and following irradiation. Here, in this research, room temperature micro-tensile specimens of non-irradiated and irradiated samples containing an Al alloy (AA 6061) bond line were tested to evaluate its strength and deformation behavior. Observations indicated that the strain rate did not affect the deformation behavior or strength and most of the micro-tensile specimens failed in a ductile mode in grains around the bond line. There was no indication that the microstructural features from the bond line affected the mechanical properties of the micro-tensile specimens. An initial examination was performed on irradiated material but further systematic studies of the effects of irradiation can be performed in the future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Radiation characterization summary of the NETL beam port 1/5 free-field environment at the 128-inch core centerline adjacent location

The characterization of the neutron, prompt gamma-ray, and delayed gamma-ray radiation fields in the University of Texas at Austin Nuclear Engineering Teaching Laboratory (NETL) TRIGA reactor for the beam port (BP) 1/5 free-field environment at the 128-inch location adjacent to the core centerline has been accomplished. NETL is being explored as an auxiliary neutron test facility for the Sandia National Laboratories radiation effects sciences research and development campaigns. The NETL reactor is a TRIGA Mark-II pulse and steady-state, above-ground pool-type reactor. NETL is intended as a university research reactor typically used to perform irradiation experiments for students and customers, radioisotope production, as well as a training reactor. Initial criticality of the NETL TRIGA reactor was achieved on March 12, 1992, making it one of the newest test reactor facilities in the US. The neutron energy spectra, uncertainties, and covariance matrices are presented as well as a neutron fluence map of the experiment area of the cavity. For an unmoderated condition, the neutron fluence at the center of BP 1/5, at the adjacent core axial centerline, is about 8.2×10 12 n/cm 2 per MJ of reactor energy. About 67% of the neutron fluence is below 1 keV and 22% above 100 keV. The 1-MeV Damage-Equivalent Silicon (DES) fluence is roughly 1.6×10 12 n/cm 2 per MJ of reactor energy.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗