Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Transient Testing”

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 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2011

Materials-Enabled High-Efficiency (MEHE) Heavy-Duty Diesel Engines

The purpose of this Cooperative Research and Development Agreement (CRADA) between UTBattelle, Inc. and Caterpillar, Inc. was to improve diesel engine efficiency by incorporating advanced materials to enable higher combustion pressures and temperatures necessary for improved combustion. The project scope also included novel materials for use in advanced components and designs associated with waste-heat recovery and other concepts for improved thermal efficiency. Caterpillar initially provided ORNL with a 2004 Tier 2 C15 ACERT diesel engine (designed for on-highway use) and two 600 hp motoring dynamometers. The first year of the CRADA effort was focused on establishing a heavy-duty experimental engine research cell. First year activities included procuring, installing and commissioning the cell infrastructure. Infrastructure components consisted of intake air handling system, water tower, exhaust handling system, and cell air conditioning. Other necessary infrastructure items included the fuel delivery system and bottled gas handling to support the analytical instrumentation. The second year of the CRADA focused on commissioning the dynamometer system to enable engine experimentation. In addition to the requirements associated with the dynamometer controller, the electrical system needed a power factor correction system to maintain continuity with the electrical grid. During the second year the engine was instrumented and baseline operated to confirm performance and commission the dynamometer. The engine performance was mapped and modeled according to requirements provided by Caterpillar. This activity was further supported by a Work-for-Others project from Caterpillar to evaluate a proprietary modeling system. A second Work-for-Others activity was performed to evaluate a novel turbocharger design. This project was highly successful and may lead to new turbocharger designs for Caterpillar heavy-duty diesel engines. During the third (and final) year of the CRADA, a novel valve material was evaluated to assess high temperature performance and durability. A series of prototype valves, composed of a unique nickel-alloy was placed in the engine head. The engine was aggressively operated using a transient test cycle for 200 hours. The valve recession was periodically measured to determine valve performance. Upon completion of the test the valves were removed and returned to Caterpillar for additional assessment. Industrial in-kind support was available throughout the project period. Review of the status and research results were carried out on a regular basis (meetings and telecons) which included direction for future work activities. A significant portion of the industrial support was in the form of information exchange and technical consultation.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2018

Summary and Assessment of Metal Fuel Capabilities in Bison

The work described in this report was performed under funding from the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. This report is issued in satisfaction of the level 2 milestone M2MS-18IN0201012 on Release Bison with Initial Metallic Fuel Performance Capability. Metallic fuel development began in Bison a few years after light water reactor development began (~ 2009). The initial development was funded by the Advanced Fuels Campaign. With recent interest from industry, the Nuclear Regulatory Commission, and the Versatile Test Reactor program, the Nuclear Energy Modeling Simulation Program has increased funding for metallic fuel development and validation in FY18. The material models with the most mature development are binary uranium-zirconium and ternary uranium-plutonium-zirconium alloys for fuel and HT9 for cladding. The first material models were incorporated from open literature sources, which include mechanical and thermal material properties that are functions of temperature, porosity, and zirconium concentration, swelling, fission gas release, zirconium diffusion, creep, and sodium coolant channel boundary conditions. These material models have been utilized to simulate ERB-II fuel pins and the results compared to measurements. The EBR-II experiment measurements mostly consist of cladding strain and zirconium redistribution. As such, comparisons to EBR-II measurements are currently limited to these two figures of merit. Simulations and comparisons to experiment measurements have also been done for transient testing of EBR- II fuel pins in TREAT, which included temperature measurements. Preliminary comparisons between Bison calculations and experiment measurements are favorable. These comparisons have also highlighted the importance of the fuel swelling and fission gas release models on cladding strain. Also, Bison metallic fuel simulations are currently being run in support of ATR experiments and VTR exploratory design calculations. Further development and evaluation of individual material models and fuel system models are planned for FY19.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Simulation of Peach Bottom turbine trip transients TT1, TT2 and TT3 with SIMULATE-3K

The Peach Bottom 2 Turbine Trip transients tests TT1, TT2 and TT3 are simulated with SIMULATE-3K. This work provides the initial basis for evaluation of the capabilities and accuracy of coupled 3D neutronics/thermal hydraulics best estimate calculations for support of safety analysis. Selected results, obtained with SIMULATE-3K, are compared to available measurements and benchmark results. The comparisons are analyzed and the application of SIMULATE-3K for best estimate analysis approach within the safety analysis methodology is discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1996

Study of a High Voltage Ion Engine Power Supply

A complete laboratory breadboard version of a ion engine power converter was built and tested. This prototype operated on a line voltage of 80-120 Vdc, and provided output ratings of 1100 V at 1.8 kW, and 250 V at 20 mA. The high-voltage (HV) output voltage rating was revised from the original value of 1350 V at the beginning of the project. The LV output was designed to hold up during a 1-A surge current lasting up to 1 second. The prototype power converter included a internal housekeeping power supply which also operated from the line input. The power consumed in housekeeping was included in the overall energy budget presented for the ion engine converter. HV and LV output voltage setpoints were commanded through potentiometers. The HV converter itself reached its highest power efficiency of slightly over 93% at low line and maximum output. This would dip below 90% at high line. The no-load (rated output voltages, zero load current) power consumption of the entire system was less than 13 W. A careful loss breakdown shows that converter losses are predominately Metal-Oxide-Semiconductor Field Effect Transistor (MOSFET) conduction losses and HV rectifier snubbing losses, with the rectifier snubbing losses becoming predominant at high line. This suggests that further improvements in power efficiency could best be obtained by either developing a rectifier that was adequately protected against voltage overshoot with less snubbing, or by developing a pre-regulator to reduced the range of line voltage on the converter. The transient testing showed the converter to be fully protected against load faults, including a direct short-circuit from the HV output to the LV output terminals. Two currents sensors were used: one to directly detect any core ratcheting on the output transformer and re-initiate a soft start, and the other to directly detect a load fault and quickly shut down the converter for load protection. The finished converter has been extensively fault tested without failure. The finished converter has been packaged suitable for use as a laboratory prototype for further testing. The finished converter is readily transportable. An article on design issues for high voltage converters for ion engines is included as an attachement.

Stuart, Thomas A.↗
DOE OSTI2017

Investigation of Treat M2 and M3 calibration experiments for benchmark analysis

The Transient Reactor Test Facility (TREAT) located at Idaho National Laboratory is a graphite moderated and reflected, air cooled core. This reactor was originally designed and operated to experimentally test fuel under extreme transient conditions. To this end, TREAT successfully performed many series of steady-state and transient tests until operations were discontinued in 1994. With the restart of the reactor planned for 2018, accurate models of previous reactor experiments are necessary to verify the ability to predict conditions for planned upcoming tests. In this work, the methodology for modeling the M2 and M3 calibration experiments (M2/M3CAL) will be described along with results from the variation of key parameters including the boron concentration. Additionally, this model will be used to demonstrate the correlation between reactor behavior and the implementation of 'nuclear graphite' thermal scattering law (TSL) data to capture the behavior of thermal neutrons in the TREAT core.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

High Burnup Transient Fission Gas Release Testing and Post-Test Examinations

The contribution of transient fission gas release (tFGR) to the burst behavior of high burnup fuel during a loss of coolant accident (LOCA) for commercial light water reactors is a complex phenomenon that is not fully understood. During the temperature ramp associated with a LOCA, there is significant additional transient fission gas release in excess of the steady state fission gas release from normal operation. An addition to the Oak Ridge National Laboratory Severe Accident Test Station (SATS) was created to explore the different conditions that impact transient fission gas release. Four tests on commercially irradiated zirconium clad uranium dioxide nuclear fuel were performed to support transient fission gas release study. These experiments and their post-test analyses are discussed. The first test on high burnup fuel revealed a LOCA relevant transient fission gas release of 5.3% for an unpressurised segment. Subsequent tests have measured transient fission gas release at different conditions that span from 4.8% to 14.4 % transient fission gas release.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Effect of natural fractures on determining closure pressure

Abstract In Utah FORGE reservoir, eight pressure transient tests (microhydraulic fracturing and DFIT™) show natural fracture/pressure-dependent leakoff. This behavior may lead to misinterpretation of the closure pressure (proxy for minimum principal stress). The closure pressures obtained from DFIT™ (or microhydraulic fracturing) testing may reflect shear failure along natural fractures or discontinuities rather than tensile failure and lead to inaccurate estimates of the minimum principal stress. In pressure tests conducted at the Utah FORGE site showed that reactivation or opening of natural fractures intersecting the wellbore and were suggested by multiple closure events in DFIT™ tests. In addition, comparisons between the pre- and post-well-tests FMI logs show that there are populations of induced fracture and reactivated natural fractures which are mostly vertical and sub-vertical. In this study, DFIT™ (or microhydraulic fracturing) test analysis and numerical simulations were used to suggest that hydraulic shearing of critically stressed natural fractures can contribute to multiple closure signatures and possibly lead to incorrect determination of the minimum principal stress. As other authors have previously advocated, the DFIT™ test analyses and numerical simulations suggest that better estimations of minimum principal stress may be derived by injecting at relatively high rate and pressure and insuring that tensile breakdown is reached. In addition, extended shut-in period is required to determine accurate reservoir characteristics and fluid flow regime.

Nadimi, S.↗
DOE OSTI2023

Microstructure and phase evolution in the U-10Zr fuel investigated by in situ TEM heating experiments

The development of U-Zr metallic nuclear fuel for fast spectrum reactors is impacted by a lack of mechanistic understanding of the fuel behavior evolution under thermal irradiation conditions, despite previous works providing substantial fuel performance data. This work uses in-situ transmission electron microscopy heating experiments to study phase and microstructural evolution in several unirradiated U-10Zr specimens during rapid heating ramps (from room temperature to 1000 °C). The starting a-U + bcc-(Zr, U) eutectic microstructure began to decompose above 600 °C. The decomposition initiated from the bcc (U,Zr) phase. Similar results were observed for all specimens even when fabricated by different routes (e.g., cold rolled or annealed). As a result, the impact of observed microstructure and phase evolutions at high temperatures on fuel fabrication and in-pile fuel transient test was also discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Distributed temperature sensing inside a 19-rod bundle

The temperature field within a model of a sodium-cooled fast reactor fuel rod bundle was measured using Ø155 μm fiber optic distributed temperature sensors (DTS). The bundle consists of 19 electrically-heated rods Ø6.3 mm and 865 mm long. Working fluids were argon and air at atmospheric pressure and Reynolds numbers up to 300. A 20 m-long DTS was threaded through Ø1 mm capillaries wound around rods as wire-wraps. The sensor generated 173 measurements along each rod at 5 mm resolution for a total of 3300 data locations. A second DTS, 58 m long, was suspended between rods to provide 9300 fluid temperature measurements at 20 mm resolution. Such data density makes it possible to construct 3D maps of the temperature field that are beyond the reach of traditional sensors such as thermocouples. This is illustrated through a series of steady-state and transient tests. As a result, the work demonstrates the feasibility of mapping temperature within the close confines of a rod bundle at resolutions suitable for validation of computational fluid dynamics codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Hyperbolic Reformulation Approach to Enable Efficient Simulation of Groundwater Flow and Reactive Transport

We apply Cattaneo's relaxation approach to the one-dimensional coupled Boussinesq groundwater flow and advection-diffusion-reaction equations, commonly used in engineering applications to simulate contaminant transport in the subsurface. The diffusion-type governing equations are reformulated as a hyperbolic system, augmented by an equation that can be interpreted as a momentum balance. The hyperbolization enables an efficient unified computation of the primary variable and its gradients, for example piezometric head and unit discharge in the Boussinesq equation. An augmented Roe scheme is used to solve the hyperbolic system. The hyperbolized system of equations is studied in a set of steady state and transient test cases with idealized geometry. These test cases confirm the equivalence of the hyperbolic system to its original formulation. The larger time step size of the hyperbolic equation is verified theoretically by means of a stability analysis and numerically in the test cases. Finally, a reach-scale application of flow and transport across a river meander is considered. This application case shows that the performance of the hyperbolic relaxation approach holds for more realistic groundwater flow and transport problems, relevant to water resources management.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2018

OPTD: Out-of-Pile Transient Database

The DOE Advanced Reactor Technology program has supported knowledge preservation efforts to recover and preserve fuel data from the US sodium cooled fast reactor (SFR) program. Efforts have focused on establishing databases of experimental data generated during the Integral Fast Reactor program (at EBR-II, FFTF, and TREAT) as well as during related out-of-pile examinations of the fuel irradiated at EBR-II. The information in these databases is essential to support further development and licensing activities for advanced fast reactor designs. OPTD, a newly-created database of out-of-pile furnace transient tests on metallic fuel, includes records of 62 tests on irradiated fuel pins conducted in furnace apparatuses constructed in the Alpha-Gamma Hot Cell Facility at ANL. Record discovery is ongoing, and at present, the database includes hundreds of documents describing these tests and related results. Access is currently available to Department of Energy researchers and can be extended to members of industry and the public as the content is approved for distribution. This report describes the out-of-pile furnace testing program, the content available in the database, and the structure of the web-accessible interface.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

FY20 Progress Report on FIPD Development and User Support

The DOE Advanced Reactor Technologies program has supported efforts to recover and preserve metallic fuel data generated throughout the past fast reactor R&D programs. Those efforts have focused on establishing databases for information from the experiments conducted during the Integral Fast Reactor program including data generated at EBR-II, FFTF, and TREAT reactors, as well as out of pile transient testing data. These databases are essential for future-licensing activities of metallic fuel based advanced fast reactors. This report describes the EBR-II Fuels Irradiation & Physics Database (FIPD) development, content updates, and user support during the current fiscal year (FY20). Future plans are also provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Thermal Spectrum Molten Salt-Fueled Reactor Reference Plant Model

This report details a fully coupled neutronics thermal hydraulics reference plant model for a thermal spectrum molten salt-fueled reactor. The Multiphysics model is developed on the Nuclear Regulatory Commission's Comprehensive Reactor Analysis Bundle available on the Idaho National Laboratory high-performance computer, which natively and seamlessly couples Griffin and Sam codes of the Multiphysics Object-Oriented Simulation Environment (MOOSE) based applications. Griffin provides the reactor physics capabilities, k-eigenvalue, delayed neutron precursor concentrations, and transient solutions for stationary and flowing fuels. Sam provides the solution of mass, momentum, and energy conservation equations of the whole system with a set of system components for one-dimensional single-phase flow. The neutronics feedback model relies primarily on fuel and moderator temperatures, fuel salt density, and delayed neutron precursors distributions. This report presents our results for the coupled steady state core and an unprotected loss of flow event. Although this model is prototypical regarding capabilities in the Comprehensive Reactor Analysis Bundle, its results are consistent with experimental data of the pump transient tests that were performed at zero power of the Molten Salt Reactor Experiment (MSRE).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Fast Reactor Database Utilization for Qualification of Metallic Fuel Forms

The DOE Advanced Reactor Technologies program has supported efforts to recover and preserve metallic fuel data generated throughout the past fast reactor R&D programs. Those efforts are currently focused on establishing databases for information from the experiments conducted during the Integral Fast Reactor program including data generated at EBR-II, FFTF, and TREAT reactors, as well as out of pile transient testing data. These databases are essential for future licensing activities of metallic fuel based advanced fast reactors. This report describes the EBR-II Fuels Irradiation & Physics Database (FIPD) development, content updates, data qualification status, and user support. Future plans are also provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Fuels Irradiation and Physics Database (FIPD) Development and Data Qualification FY24 Updates

The DOE-NE Advanced Reactor Technologies program has supported efforts to recover and preserve metallic fuel data generated throughout the past fast reactor R&D programs. Those efforts are currently focused on establishing databases for information from the experiments conducted during the Integral Fast Reactor program including data generated at EBR-II, FFTF, and TREAT reactors, as well as out of pile transient testing data. These databases are essential for future-licensing activities of metallic fuel based advanced fast reactors. This report describes the Metallic Fuels Irradiation & Physics Database (FIPD) development, content updates, data qualification status, and user support in FY24. Future plans are also provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Baseband Digital Network Analyzer Upgrade for LLRF Controllers

Digital Network Analyzers (DNA) have been implemented in many Low-Level Radio Frequency (LLRF) systems, notably NSLS-II and CERN, to help tune feedback loops. DNA characterizes feedback loops by measuring the frequency-dependent magnitude and phase transfer functions. It enables the measurement of open loop gains, gain/phase margins, and loop delays to help fine-tune feedback loops. An FPGA-based DNA has been developed and integrated into the current Relativistic Heavy Ion Collider (RHIC) LLRF infrastructure. Its performance has been tested with an implementation of one-turn delay feedback (OTFB) on the bench to maximize gain and stability. The DNA has been used to characterize a RHIC 28 MHz cavity in a RHIC Accelerator Physics Experiment (APEX) to test transient beam loading compensation strategies.

43 PARTICLE ACCELERATORS↗
DOE OSTI2025

Griffin Capability Improvements in Support of Ex-core Deep-Penetration Problems

Advanced reactor designs, especially portable reactors that are designed to be located closer to humans and operate autonomously, require the ability to accurately compute the ex-core neutron and gamma flux solutions in terms of shielding design optimization to reduce dose rates at the vessel boundary and detector signal prediction to drive the reactor control system. The Nuclear Energy Advanced Modeling and Simulation program has prioritized improvements to the Griffin discrete ordinates (SN) solver for deep-penetration problems in fiscal year 2025. Significant advancements have been made to the Griffin methodologies for solving ex-core deep-penetration problems for steady-state, fixed-source and transient calculations. This work presents the methodology improvements as well as a comprehensive demonstration with a Transient Test Reactor model and measurements.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Loss of Flow Conditions in a Modern Pool-type SFR and In-Pile Experiment Design

The metallic fuel safety performance under unprotected design-basis transients is a key consideration for the deployment of advanced sodium fast reactors (SFRs). Reliable data are needed to validate advanced safety codes, reduce uncertainty in cladding failure thresholds, and strengthen confidence in licensing approaches. To address this need, this report develops blueprints for a conceptual sodium loss-of-flow (LOF) experiment in the Mk-IIIR loop at the Transient Reactor Test Facility (TREAT), providing the technical foundation for future integral testing.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗