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 397 records · Page 22
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

MCNP Analysis fo the Aqua-SETH Experiment in TREAT

The versatile Transient Reactor Test (TREAT) Facility’s capability to deliver customized transient nuclear power excursions gives researchers and developers a multitude of specialized testing options. In conjunction with reactorprescribed delivery of neutron flux, experiment design requires specimen packaging to meet TREAT safety containment requirements and experimenter test objectives. These objectives may range from the very simple separate effects tests (generally dry capsule) to the more complex integral scale testing wherein prototypic boundary or specimen reactor environment conditions are created. The historic MARK-III sodium flow loop that achieved thermalhydraulic prototypic test conditions serves as an example of the more complex experiment design package.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Report on RIA Relevant Modified Burst Testing of ATF Cladding Materials

The mechanical performance of accident-tolerant fuel (ATF) cladding candidates in light-water reactors (LWRs) must be similar to or better than that of current conventional nuclear fuel claddings to reduce dose to the public and ensure a that the core coolable geometry is maintained during a postulated reactivity-initiated accident (RIA) in light-water reactors (LWRs). During an RIA event, the rapid thermal expansion of nuclear fuel can deform the cladding once the fuel–cladding gap closes. In some cases, the pellet–cladding mechanical interaction (PCMI) can induce mechanical failure in ATF candidates. Thus, the mechanical response of ATF cladding must be investigated by mimicking the conditions of RIA and potentially performing Transient Reactor Test (TREAT) experiments to establish or verify the safety envelope. The work presented in this report investigated the failure behavior of as-received, hydrided, and chromium-coated (Cr-coated) Zircaloy-4 (Zry-4) cladding tube under strain-driven mechanical conditions, mimicking postulated RIA loading conditions. Mechanical testing was performed at 300°C via modified burst test (MBT) equipment with pulse width control previously developed under the Department of Energy’s (DOE’s) Advanced Fuel Campaign (AFC). The mechanical strains were determined using 2D digital image correlation (DIC) techniques. The base Zry-4 acquired by Cameco Inc. was in stress-relieved annealed (SRA) condition. Because of the observed large deformation of the cladding tubes, the failure strain definition was updated for the MBT, which can also be applied to other tube tests where significant bulging (out-of-plane deformation) is present. The failure strain was determined to be affected by the speed of the test or the RIA event. As the RIA-like event duration decreased from 75 to 15 ms, the failure strain decreased 5, 7 and 1% for as-received, hydrided, and Cr-coated specimens, respectively. Fractography on the Cr-coated specimens indicated the presence of two failure mechanisms: (i) crenulation at the outer surface of the coating due to the tensile tractional forces along with coatings grain microstructure and (ii) formation of critical defect at the coating/cladding interface that initiated coating rupture after severe plastic deformation of the Zry- 4 substrate. Based on the MBT results and fractography observations, performing mechanical property testing at high strain rates—in particular on Cr-coated tubes—and semi-integral TREAT experiments are required future efforts to ensure ATF cladding performance during transients. The testing recommended would also inform the development of long-term generalized cladding technologies.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Effects of Boron and Graphite Uncertainty in Fuel for TREAT Simulations

As a result of the accident at Fukushima, the United States has made a concerted effort to develop fuel with enhanced accident tolerance for the current fleet of reactors. In 2012 Congress laid out a schedule for the development of accident tolerant fuel (ATF), the goal of which is to test in a commercial reactor by 2022. Idaho National Lab's (INL) Transient Reactor Test Facility (TREAT) is one of the locations that will be performing ATF testing and is scheduled for restart in 2018. TREAT was built to subject fuel to conditions ranging from mild transients to severe reactor accidents. However, typically a number of pre-test calibrations are required before evaluation of fuel can begin, which is costly and time consuming. Advanced modeling techniques and current computational capacity make full core TREAT simulations possible, with the goal of such simulations to understand the pre-test core and minimize the number of required calibrations. But, in order to simulate TREAT with a high degree of precision the reactor materials and geometry must also be modeled with a high degree of precision. This paper examines how uncertainty in the reported values of boron and graphite have an effect on simulations of TREAT.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1993

Evaluation of Conservatism in Low Frequency Vibration Test Control

Aerospace structures are typically qualified for low frequency flight environments by using a swept-sine vibration test method. However, this ground testing has a tendency to produce substantial overtesting at certain frequencies. This paper presents an evaluation of the response conservatism present with a typical swept-sine and two transient vibration tests when applied to a typical spacecraft component. The absolute conservatism between a typical launch transient response and the test environment responses is measured using alternative characterizations of transient vibrations previously used in shock testing. The characterizations used include the Shock Intensity Spectrum, Shock Response Spectrum, Acceleration Root Mean Square in both frequency and time domains, and Ranked Peaks. Control of the test/flight conservatism is shown to be possible through the overtest factor (OTF) parameter.

Conservatism↗
DOE OSTI2026

Refinement and demonstration of a coupled BISON-Griffin workflow for designing targeted TRISO transient experiments in TREAT

The U.S. nuclear industry is expected to deploy tristructural isotropic (TRISO) particle fuel technologies for commercial reactors within the next decade. In previous work, we defined a preliminary transient design space for TRISO fuels, identified potential gaps in the available data, and began to develop multiphysics modeling tools that could be applied to design targeted Transient Reactor Test Facility (TREAT) experiments to fill these gaps. Here, this work builds on that foundation by (1) updating BISON fuel performance and Griffin reactor physics models to reflect the current TREAT experiment tube and capsule designs, (2) coupling the codes to improve the accuracy and usability of the transient design analyses, and (3) demonstrating their use over an expanded design space that includes fuel burnup. The simulated mechanical responses of the TRISO particles were complex functions of fission product accumulation, fission gas release, and irradiation-induced dimensional change in the pyrolytic carbon layers. The predicted tangential stresses in the particles' silicon carbide layers were least compressive for preheated tests involving fresh fuels but remained compressive throughout the ranges of temperature, heat rate, and burnup considered in this work. Finally, comparisons between the potential TREAT transients and historical test reactor irradiations showed that the TREAT tests would produce significantly lower average energy deposition rates, yielding less severe transients with greater relevance to near-term commercial applications. The use of these predictive capabilities has the potential to increase the value of each test, improving the overall efficiency and cost-effectiveness of transient testing for TRISO and other advanced fuels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

On the origin of permanent performance loss of anion exchange membrane fuel cells: Electrochemical oxidation of phenyl group

The durability of alkaline anion exchange membrane fuel cells (AEMFCs) is a critical requirement for implementing this technology in cost-effective energy conversion. However, the lifetime of current AEMFCs is short, ca. < 1,000 h and the cause of the performance loss over time is poorly understood. Here, we investigate the origin of AEMFC unrecoverable performance loss. We diagnosed the recoverable performance loss induced by carbonation and anode flooding during an extended-term test with transient operating conditions. Multiple extended-term tests under current densities and the electrochemical and analytical analyses suggest that the most prominent permanent performance loss of AEMFCs occurs by phenol formation as a result of electrochemical phenyl oxidation of cathode ionomeric binder. The produced phenol is acidic and neutralizes the quaternary ammonium group to lower the local pH. Finally, the impact of this degradation pathway is discussed to explain the reported lifetime performance of current AEMFCs.

08 HYDROGEN↗
DOE OSTI2016

Effects of Coolant Void Condition on Melt Fragmentation using Simulants in Initial Phase of Metal-fueled SFR Severe Accident

Metal fuel has inherent safety characteristics, especially in severe accident conditions. However, the metal-fueled SFR severe accident scenarios need to be categorized since overall severe accident mitigation strategies have not been prepared. In addition, these mitigation strategies should be classified in specific phases. In an initial phase of metal-fueled SFR severe accident scenarios, fuel melting, and cladding failures would occur. After the occurrence of the cladding breach, the molten metal fuel would be ejected into a coolant channel due to fission gas in fuel cavity or rapid evaporation of sodium bond. This is called Ex-pin phenomena. There are few studies considering the Ex-pin phenomena of metal fuel. Related representative studies are transient reactor test facility (TREAT) M series tests. The TREAT M series tests provided fundamental metal fuel performance in transient condition. They informed partial experimental data such as the amount of melt ejected into coolant channel after the cladding failure. However, these experimental data did not examine the melt behavior during the Ex-pin phenomena. In metal-fueled SFR, it is a known fact that the negative reactivity feedback occurs in the initial phase of severe accident. In an unprotected loss-of-flow (ULOF) accident, the void-reactivity feedback may be positive depending on where in the reactor core the sodium voiding occurs. However, molten metal fuel would be better dispersed upward compared to those of an oxide fuel and significant negative reactivity feedback could be induced through such behaviors. Thus, a key safety strategy to mitigate the accident progression at the initial phase of severe accident is to discharge melt upwards out of the core. In this paper, the upward discharge of melt is defined as melt levitation. Such melt levitation could ensure early termination of severe accident. There could be different melting behaviors inside the pin depending on reactor types or the type of fuel. In case of pressurized water reactor (PWR) oxide fuel-water system, the rapid power excursion would initiate melting of fuel at the surface of the pellets. In this paper, the ULOF accident for the metal-fueled SFR is considered as an accident scenario. To induce the melt levitation, the driving force should be clarified. Heo et al. observed a built-up vapor pressure was suggested as one of the driving forces for the upward dispersion of the molten materials. The following factors also have the potential to induce the melt levitation; the pressure induced by fission gas in the cavity or rapid boiling of sodium bond, natural circulation flow of the sodium, the sodium vapor flow generated by the difference between inlet and outlet plenum in the reactor core, the built-up vapor pressure due to the fuel-coolant interaction etc. These factors are balanced each other. They should be examined through the Ex-pin experiments. Before conducting a parametric study for each factor, the general behavior of the molten metal fuel should be investigated in the coolant channel. Since rapid sodium coolant boiling could occur in the initial phase of severe accident scenarios, the molten fuel would be ejected and fragmented in voided channel. Unlike previous studies concerning an un-voided channel, differences in melt behavior in the voided channel are expected. In present study, molten material behavior was evaluated to reveal their characteristics in voided channel. The Ex-pin experiments were performed in both un-voided and voided channel, thus such behavior differences were explicitly observed. In the lab-scale experiments, the un-voided and voided channel were reflected by coolant-filled and air-filled channel, respectively. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Post-irradiation examination of the Sirius-1 nuclear thermal propulsion fuel test

Nuclear Thermal Propulsion (NTP) systems hold promise in reducing transit times for exploration of the solar system by both crewed and uncrewed missions. NTP systems currently under investigation include a once-through high-temperature gas-cooled fission reactor to provide thermal energy to heat the coolant, which also serves as the propellant. The fuel system consists of angular UN fuel particles dispersed in a matrix of W/Re, creating a ceramic and metallic composite or cermet, which has been irradiated in Idaho National Laboratory's (INL) Transient Reactor Test Facility (TREAT). This enabled evaluation of these materials under representative nuclear heating rates (~95 K/s) and peak surface temperatures (~2527 K). Further, these surface conditions were indicative of achievement of the target peak internal temperature of approximately 2850 K. These tests named Sirius-1 (UN–W/Re), have been irradiated and this paper will present post-irradiation examination results. The Sirius-1 test produced minor, stable cracks in the fuel specimen and spalling of surface material. Volatilized uranium ‘soot’ was found deposited on the inner wall of the irradiation capsule indicating loss of some fissile material from the fuel specimen. Spalling from the surfaces was also noted upon visual inspection. Uranium diffusion from the fuel particles resulted in the formation of U/Re phases and the production of a laminar microstructure on the edge of the fuel system. Overall, the specimen was stable and remained within a coolable geometry upon conclusion of multiple thermal cycles to prototypical operating temperatures.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

COR-0011 Rev 6 MARVEL Project Code of Record

This Code of Record identifies the codes, standards, and procedures necessary to design, develop, construct, and startup the Microreactor Applications Research Validation and Evaluation (MARVEL) Project at the Materials and Fuel Complex (MFC) Transient Reactor Test (TREAT) Facility and the Idaho National Laboratory (INL). The MARVEL Project is an INL test microreactor funded by the United States Department of Energy (DOE) via the Microreactor Program (MRP). The goal of the project is to establish an operational nuclear applications test bed that can generate combined heat and power to enable integration and R&D with end-user technologies, as well as allow microreactor technologists to test next-generation control systems. The microreactor is a thermal reactor utilizing Uranium Zirconium Hydride (UZrH) fuel with review and authorization by the Department of Energy Idaho Operations Office (DOE-ID) for National Environmental Policy Act (NEPA) compliance, safety review, and supplemental readiness assessments for startup and operation. To enable rapid deployment, the MARVEL reactor will reside in the Transient Reactor Test (TREAT) Facility and utilize the existing operating Category B reactor facility, approved facility safety basis, operating crews, and recent re-start experience.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
NASA NTRS1988

Transient performance evaluation of an integrated heat pipe-thermal storage system

Transient performance tests of an integrated heat pipe-thermal storage system have been conducted. This system was developed as a part of an Organic Rankine Cycle-Solar Dynamic Power System receiver for future power systems. The integrated system consists of potassium heat pipe elements that incorporate thermal energy storage canisters within the vapor space and an organic fluid (toluene) heater tube used as the condenser region of the heat pipe. The transient performance tests determined the operating characteristics and power input limits of the integrated heat pipe-thermal storage unit under conditions corresponding to re-acquisition of the sun during emergence from eclipse conditions and to the initial start-up of the solar dynamic power system. The tests demonstrated that the heat pipe-thermal storage element is not limited under conditions corresponding to emergence from eclipse during normal orbital operations and the heat pipe will successfully start-up from the frozen condition with full input power at the onset. Details of the test procedures and results of the tests are presented in this paper.

Keddy, E.↗
DOE OSTI2019

Second Annual Progress Report on Ultrasonic Sensors for TREAT Fuel Condition Measurement and Monitoring

Fuel deformation (both axial and radial) information provides insights into fuel performance and failure limits. Traditionally, fuel deformation has been quantified using post-irradiation examinations due to the lack of an appropriate sensor for monitoring deformation in situ. Recent developments in sensor and instrumentation design have led to the use of linear variable differential transformers (LVDT) for monitoring axial strain in fuel during irradiation. Size and operational constraints limit the use of LVDT sensors for radial deformation monitoring, and alternatives are needed. One possible alternative to LVDT sensors is an ultrasonic sensor. Ultrasonic sensors monitor the interaction of high-frequency stress waves with the test object. The resulting response may be analyzed to determine the properties of the material, including stress, strain, microstructure, presence of defects, and the elastic constants. While ultrasonic measurements have been successfully applied for monitoring strain in a number of applications, their use for in-core measurements has been limited. Largely, the limitations are due to the challenges associated with designing sensors that can operate under typical in-core conditions. The overall objective of this work is to design an ultrasonic sensor capable of rapid in-situ measurements of fuel rodlet dimensional changes during TREAT transient irradiation testing. This report summarizes the activities and research findings through the end of Year 2 of this three-year project.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Real-World 20-Hz IBR Subsynchronous Oscillations: Signatures and Mechanism Analysis

This article provides analyses of subsynchronous oscillations associated with solar photovoltaics (PVs) observed in real world. All have oscillation frequencies at approximately 20 Hz in voltage and current magnitudes. On the other hand, in phase voltage and current, the oscillations show very different signatures. Through constructing simplified linear models and conducting sensitivity analysis, we are able to associate dynamic phenomena with the underlying mechanism and speculate the possible causes of the 20-Hz oscillations. The analysis results have all been validated via simulation results from electromagnetic transient (EMT) test beds constructed in MATLAB/Simscape.

inverter-based resources↗
DOE OSTI2015

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the power coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.

Connaway, H. M.↗
DOE OSTI2023

A non-fueled nuclear-heated rod for in-pile transient boiling studies

Separate-effects boiling experiments have recently been conducted in the Transient Reactor Test Facility at Idaho National Laboratory to investigate transient heating and irradiation effects on cladding-to-coolant heat transfer. Specifically, transient critical heat flux (CHF) remains an important area of research, and better understanding of this phenomenon has potential for improving predictive models related to operational and safety limits. Consequently, this knowledge is expected to improve efficiency of light-water reactor operations. A novel borated nuclear-heated rodlet (BNHR) was designed to enable observation of transient cladding-to-coolant heat transfer phenomena. The final BNHR design takes a surrogate approach, wherein nuclear heating is induced by 10 B(n, α) reactions rather than derived from fissions in a fueled specimen. The structure of the BNHR consists of a hollowed out borated (B nat ~ 2.05 wt %) stainless steel tube with an hourglass-shaped outer surface, capped at both ends with non-borated stainless steel. This geometry allows for inner-rodlet instrumentation and generation of the highest nuclear heating rates near the center of the rodlet to ensure onset of boiling near instrumentation for real-time observation. A novel approach to measuring the nuclear energy deposition rate in the BNHR separate and apart from the influence of the coolant, termed the n-a thermometer, is also detailed in this paper. This device has demonstrated excellent repeatability, and measurements indicate predictive modeling results for energy deposition in the BNHR rod agree within a 10% margin of the experiment measurements. In conclusion, these results give confidence that the BNHR design has successfully met experiment objectives.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

TREAT HODOSCOPE – A Predictive Simulation Model

The Transient Reactor Test Facility (TREAT) reactor is being restarted after over two decades of standby operation. TREAT is designed to test the time-dependent behavior of nuclear fuel during simulated reactor accidents. One of the principle data collection systems of TREAT is the fast neutron hodoscope which is used to observe the motion of the fuel test specimen as it is being damaged, deformed, or melted over the course of the power transient. The hodoscope is a grid of ex-core collimator channels paired with a bank of fast neutron sensitive radiation detectors. These detectors have historically been calibrated using empirical methods utilizing measurement data from smaller power pulses. Given modern computational capabilities this work sets out to establish a predictive computational capability of the fast neutron flux field exiting the hodoscope to support these calibrations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

Accelerated Climate Model for Energy (ACME): Ocean and Sea-Ice Processes

The primary goal of this project was to understand the role of the ocean and sea-ice components in the evolution of the climate in both a high-resolution fully-coupled 1850 preindustrial control and present day transient climate simulations run using the Energy Exascale Earth System Model (E3SM), formerly known as the Accelerated Climate Model for Energy (ACME). Two distinct approaches to the initialization of the present day transients were tested: in the first, an ensemble of three transients used atmospheric reanalysis forced 0.1-degree ocean/sea-ice states, while a fourth transient was branched off the preindustrial control. Biases from the control were inherited by its successor transient, in particular, sea surface temperatures (SSTs) at high latitudes were too warm and too much northward meridional heat transport occurred in the North Atlantic Ocean. SSTs in the Northern Hemisphere and Atlantic meridional heat transport increased in the transients initialized from the ocean/sea-ice simulation over a 40-50 year period to largely resemble the transient branched off the control simulation. Excessive absorbed solar radiation (ASR) is considered to be responsible for the SST biases as well as the formation of mode waters in lighter density classes than in reality both in the Southern Ocean and the North Atlantic Ocean subtropical gyre. These mode waters ventilate the thermocline waters to the north of their formation sites explaining the excessive Atlantic meridional heat transport and some of the warming in the high-latitude North Atlantic Ocean after multiple decades.

54 ENVIRONMENTAL SCIENCES↗
NASA NTRS2004

Silicon Carbide Diodes Performance Characterization at High Temperatures

NASA Glenn Research center's Electrical Systems Development branch is working to demonstrate and test the advantages of Silicon Carbide (SiC) devices in actual power electronics applications. The first step in this pursuit is to obtain commercially available SiC Schottky diodes and to individually test them under both static and dynamic conditions, and then compare them with current state of the art silicon Schottky and ultra fast p-n diodes of similar voltage and current ratings. This presentation covers the results of electrical tests performed at NASA Glenn. Steady state forward and reverse current-volt (I-V) curves were generated for each device to compare performance and to measure their forward voltage drop at rated current, as well as the reverse leakage current at rated voltage. In addition, the devices were individually connected as freewheeling diodes in a Buck (step down) DC to DC converter to test their reverse recovery characteristics and compare their transient performance in a typical converter application. Both static and transient characterization tests were performed at temperatures ranging from 25 C to 300 C, in order to test and demonstrate the advantages of SiC over Silicon at high temperatures.

Lebron-Velilla, Ramon C.↗