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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 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Scalable waveguide design for three-level operation in Neodymium doped fiber laser

We have constructed a double clad neodymium doped fiber laser operating on the three-level 4 F 3/2 → 4 I 9/2 transition. The laser has produced 11.5 W at 925 nm with 55% slope efficiency when pumped at 808 nm, comparable to the best previous results for a double-clad fiber configuration on this transition. Higher power pumping with both 808 nm and 880 nm sources resulted in an output of 27 W, albeit at lower slope efficiency. In both cases, output power was limited by available pump, indicating the potential for further power scaling. To suppress the stronger four-level 4 F 3/2 → 4 I 11/2 transition we developed a waveguide that provides spectral filtering distributed along the length of the fiber, based on an all-solid micro-structured optical fiber design, with resonant inclusions creating a leakage path to the cladding. Furthermore, the waveguide supports large mode areas and provides strong suppression at selectable wavelength bands, thus easing the restrictions on core and cladding sizes that limited power scaling of previous approaches.

42 ENGINEERING↗
DOE OSTI2022

Analyzing the effect of pressure on the properties of point defects in $\gamma\mathrm{U–Mo}$ through atomistic simulations

Uranium–molybdenum (U–Mo) alloys in monolithic fuel foil are the primary candidate for the conversion of high-performance research reactors in the USA. Monolithic fuel is utilized in a plate-type design with a zirconium diffusion barrier and aluminum cladding. These fuel types are unique in that they contain no plenum for the release of fission gases, which, in conjunction with the aluminum cladding, can lead to large stress states within the fuel. The nature of how fundamental processes of radiation damage, including the evolution of point defects, under such stresses occur is unknown. In this work, we present molecular dynamics simulations of the formation energy of point defects under applied stress. We report this work will allow for the implementation of stress-dependent microstructural evolution models of nuclear fuels, including those for both fission gas bubble growth and creep, which are critical to ensure the stable and predictable behavior of research reactor fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Measure Guideline: Deep Energy Enclosure Retrofit for Double-Stud Walls

This Measure Guideline describes a deep energy enclosure retrofit solution that provides insulation to the interior of the wall assembly with the use of a double-stud wall. The guide describes two approaches to retrofitting the existing walls—one that involves replacing the existing cladding and the other that leaves the cladding in place. This guideline also covers the design principles related to the use of various insulation types and provides strategies and procedures for implementing the double-stud wall retrofit. It also includes an evaluation of important moisture-related and indoor air quality measures that need to be implemented to achieve a durable high-performance wall.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2015

Uncertainty Quantification and Management for Multi-scale Nuclear Materials Modeling

Understanding and improving microstructural mechanical stability in metals and alloys is central to the development of high strength and high ductility materials for cladding and cores structures in advanced fast reactors. Design and enhancement of radiation-induced damage tolerant alloys are facilitated by better understanding the connection of various unit processes to collective responses in a multiscale model chain, including: dislocation nucleation, absorption and desorption at interfaces; vacancy production, radiation-induced segregation of Cr and Ni at defect clusters (point defect sinks) in BCC Fe-Cr ferritic/martensitic steels; investigation of interaction of interstitials and vacancies with impurities (V, Nb, Ta, Mo, W, Al, Si, P, S); time evolution of swelling (cluster growth) phenomena of irradiated materials; and energetics and kinetics of dislocation bypass of defects formed by interstitial clustering and formation of prismatic loops, informing statistical models of continuum character with regard to processes of dislocation glide, vacancy agglomeration and swelling, climb and cross slip.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Initial assessment of clad/base metal systems for MSR

In support of the development of new design and construction rules and analysis methods for cladded structure for use under the anticipated service lifetime and cyclic service conditions in Molten Salt Reactors (MSR), research work has been initiated at ORNL for assessment of physical, metallurgical and mechanical properties of clad/base metal systems. This report summarizes the results for the nickel cladded SS316H system in FY18.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Second Target Station High-Fidelity Target Activation Comparison

The development of the Second Target Station (STS) target system at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory (ORNL) is well underway. The target system at STS consists of a rotating target disk that contains 21 segments of tungsten clad in tantalum clad in steel. A key aspect of the design of the target system is to account for the delayed heating and material damage caused by the delayed dose from decaying radionuclides. These radionuclides are a product of either spallation reactions or transmutation of the nuclei in the target system. These radionuclides build up in the target system components over the lifetime of the facility, and the radiation that is emitted can deposit energy in the components causing significant component heating and material damage. Monte Carlo N-Particle (MCNP) Version 6.2 transports the various particle species and calculates the spallation products and neutron fluxes throughout the target system. These spallation products and neutron fluxes along with the material definition of each component are relayed to the CINDER2008 transmutation code to calculate the radionuclide inventories and the corresponding decay gamma emission spectra. MCNP6.2 coupled with CINDER2008 is the computational method-of-choice for the analysis discussed in the following sections of this report. The analysis focuses on validating major assumptions in calculating the radionuclide inventory in the STS target system: all of the target segments are fresh, unirradiated material when the protons are incident on the segment, the average of the 21 segments of the target is sufficient to represent a single segment, and that averaging the proton pulse structure over time does not significantly affect the radionuclide inventory. The position-averaged, high-fidelity, and single-tally computational methods are used to validate the assumptions and provide a point of comparison to evaluate how the assumptions impact the radionuclide inventories. A more detailed explanation of the three computational methods is provided in Section 2. The position-averaged and single-tally methods are less computationally expensive when compared with the high-fidelity method where 54,000 individual calculations are needed to calculate 1 hr of STS operation. Section 3 details the comparison of the three methods to show that the assumptions made in the position-averaged method do not significantly impact the radionuclide inventory after 1 hr of operation. The discussions and results in this report are for 1 hr of operation. Due to the computational cost associated with calculating the transmutation and activation using the high-fidelity method, only 1 hr of operation has been calculated. The discrepancies observed after 1 hr of operation are not extrapolated out to longer operational times, and this report does not address how the discrepancies between the computational methods may manifest for longer operational periods.

43 PARTICLE ACCELERATORS↗
DOE OSTI2017

IN-PILE MEASUREMENTS OF FUEL ROD DIMENSIONAL CHANGES UTILIZING THE TEST REACTOR LOOP PRESSURE FOR MOTION

Different types of materials are being considered for fuel, cladding, and structures in existing and advanced nuclear reactors. These materials can undergo significant dimensional and physical changes during irradiation. Currently in the US, such changes are measured with the ‘cook and look’ method (i.e. repeatedly irradiating a specimen for a specified period of time and then removing it from the reactor for evaluation). The time and labor to remove, examine, and return irradiated samples for each measurement makes this approach expensive. In addition, such techniques provide limited data, and handling may disturb the phenomena of interest. Therefore, in-pile detection of changes in geometry is sorely needed to understand real-time behavior during irradiation testing of fuels and materials in high flux US Material and Test Reactors (MTRs). This paper presents development results of an advanced Linear Variable Differential Transformer (LVDT)-based test rig capable of detecting real-time changes in diameter or length of fuel rods or other material samples during irradiation in US MTRs. This test rig is being developed at the Idaho National Laboratory (INL), and will provide experimenters with a unique capability to measure dimensional changes associated with fuel and clad swelling, pellet-clad interaction, and crud buildup. This novel LVDT design uses the MTR’s loop pressure to drive the test specimen at a constant rate. This leads to a simplified construction, that is, the use of fewer components than other diameter gauge test rigs. This simplified design only requires three pressure boundary penetrations. Further, a key component of this test rig is a leak tight hydraulic ram capable of actuating under low differential pressure, and is able to survive the high irradiation tests at Pressurized Water Reactor (PWR) conditions. The sensor has been proven on a benchtop scale, and shows repeatability and reliability. Measurement accuracies of ±3 µm for diameter and ±90 µm for length have been achieved.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2020

Neutronic Consideration of TREAT Facility Fuel SiC Recladding

The Transient Reactor Test (TREAT) Facility is an air-cooled reactor composed of graphite-urania fuel encapsulated in zirconium alloy canisters and surrounded by ~2 ft (~61 cm) of nuclear grade graphite. The primary role of TREAT is to provide in-pile nuclear-heated safety research via transient testing [1]. Benchmark efforts have demonstrated sensitivity of TREAT neutronics models to graphite cross section data and key neutron poisons found in the fuel assemblies [2]. During fabrication of TREAT fuel assemblies, a small concentration of boron was inadvertently introduced with variable concentration into the aggregate of fuel blocks prepared for TREAT use. The concentration of boron, and its respective uncertainty, represent the most sensitive component in a neutronics study; unfortunately, nothing can be done to reduce or otherwise mitigate its impact. Hydrogen in the fuel blocks reduces the mean free path of the core and its true quantity and uncertainty is approximated. The hydrogen remains from pitch binder used in the fuel fabrication process and was not completely released due to incomplete fuel graphitization. Destructive fuel testing is necessary to quantify the actual hydrogen content and to improve benchmark model quality. A third neutron poison of some significance is the hafnium impurity in the Zircaloy-3 (Zry-3) assembly cladding. The quantity of hafnium is expected to be within nominal manufacturing limits for nuclear grade zircaloy. Recladding the fuel assemblies with material that is less neutron absorbing is expected to benefit reactor neutron economy. Moreover, TREAT’s total transient energy is restricted based on the rate of Zry-3 oxidation in TREAT’s air coolant environment. Hence, cladding with more benign air reaction kinetics could dramatically uprate the facility’s experimental capability. Silicon carbide (SiC) has been proposed as a candidate material for recladding TREAT assemblies, as SiC/SiC demonstrate favorable properties within nuclear environments [3]. The current operational limits of TREAT are restricted by the oxidation of Zry-3. Change of the cladding to SiC could feasibly double the current temperature limitation. Recladding opportunities allow for previously damaged fuel assemblies to be returned to service, as Zry-3 is not commercially available. The initial effort required to consider proceeding with any recladding effort is to neutronically characterize key reactor parameters while replacing Zry-3 with SiC. A characterization summary, with respective results, is provided herein.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

ASNF Engineering Scale Drying Experiment Acceptance Criteria for Fabricated Items

This document specifies the acceptance criteria for items fabricated by Holtec International under the scope of INL Contract # 215725, consistent with the parameters described in INL/EXT-19-56017, Aluminum-clad Spent Nuclear Fuel Engineering Scale Drying Experiment Design. Verification is confirmed with initials (or the need for remediation may be identified) along with the date of inspection for each requirement. Verification may be performed by independent proxy (University of South Carolina participant, under INL Contract Number 207046) or by an INL employee, remotely by camera images.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2020

ASNF Engineering Scale Drying Experiment System Operability Functional Requirements

This checklist specifies the functional requirements for the drying system installed at Holtec International under the scope of INL Contract # 215725, in collaboration with University of South Carolina under INL Contract Number 207046, consistent with the parameters described in INL/EXT-19-56017, Aluminum-clad Spent Nuclear Fuel Engineering Scale Drying Experiment Design. Completion of the ASNF Engineering Scale Drying Experiment Acceptance Criteria for Fabricated Items, INL/MIS-20-58551, is a prerequisite.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Development of Accelerated Steady-state Test Capsule Experiments to Replicate EBR-II Fuel Behavior Using BISON Fuel Performance Analysis

Here in this work, BISON fuel performance calculations were performed to predict the fuel behavior of accelerated burnup U-Pu-Zr fuel, with temperature operation conditions of the fuel and the cladding mirroring conditions within EBR-II fuel pins. The temperature operating conditions within the FAST accelerated burnup rods were aimed at replicating EBR-II X447/X447A fuel surface and inner cladding surface temperatures. Due to the FAST capsule design, these temperatures can be replicated with fission rate densities being significantly increased. fuel performance modeling has not been assessed for novel experiments such as accelerated burnup utilizing the FAST capsule within ATR. This is an important step in understanding accelerated irradiation methods as many performance models are empirical models conforming to the results of PIE but do not always include physical models that would represent the changes in irradiation tests.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

A study of transient FGR by integral LOCA tests

A study of transient fission gas release (FGR) was performed by means of two out-of-pile LOCA tests in a hot cell facility. Irradiated fuel rod segments with lengths of roughly 80 mm and burnups of 64 MWd/kgU were subjected to simulated water-cooled reactor LOCA transients in an integral LOCA test device. The segments were heated to a maximum temperature of 1000 Celsius degrees and the increase in rod inner pressure resulting from fission gas release and increased gas temperature during the transient was measured. The first release of gas occurred at temperatures close to 450 Celsius degrees, which suggests a burst release mechanism that may be linked to the onset of grain boundary fracturing. At the end of the simulated transients, gas samples were taken from the fuel rod segments for study by gas mass spectroscopy. The transient fission gas release was measured and amount to roughly 13 % of the calculated pre-test rod inventory of Kr and Xe. Isotopic ratios of the released gas indicated that transient FGR originated from the whole pellet and not only the pellet periphery. The tested segments were also studied by light-optical microscopy to determine pellet porosity, pellet crack pattern and fuel fragmentation. The extent of fine fuel fragmentation was found to be small and limited to the periphery of the fuel. The determined amount of transient fission gas release can be added to the database of fuel rod design codes to more accurately describe the LOCA cladding strains and burst temperatures in the future.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Postirradiation Examination of the ATF-1 Experiments - 2019 Status

A collaborative effort of the Advanced Fuels Campaign together with industry consortia is focused on the development of enhanced accident tolerant fuels for Light Water Reactors (LWRs). The scoping studies, referred to as ATF-1 irradiations, are being performed in the Advanced Test Reactor of Idaho National Laboratory using drop-in style irradiations. The Post-Irradiation Examinations (PIEs) of ATF-1 began in 2017 and continued in 2019. This report describes the results of examinations on ATF-1 irradiations designed to test the performance of advanced iron-based cladding, including compatibility of FeCrAl with UO2, and ATF-1 irradiation of U3Si5. Two rodlets with commercial iron-based cladding and UO2 pellets have been irradiated to low burnup (i.e., < 20 GWd/tHM). The first rodlet had an Alloy33 cladding, while the other had a Kanthal APMT® cladding. The performed PIEs were focused on the determination of fuel microstructure, evaluation of fuel-cladding interaction and irradiation-induced variations of cladding mechanical properties. For both rodlets, no significant changes in the cladding hardness were measured, and the cladding hoop strain remained limited. Gamma tomography revealed no Cs axial migration as expected for LWR fuel, but significant cesium radial migration, which was induced by thermal migration consequent to high irradiation temperatures. Microstructure analyses revealed formation of secondary phases along the pellet rim of samples from both rodlets. No wastage or extensive chemical interaction between the fuel and the internal wall of the cladding was observed, suggesting that the secondary phases accumulated at the pellet rim might be due to formation of Cs-U-O compounds. However, further advanced PIEs are necessary to finally confirm this hypothesis. Optical microscopy analysis of the rodlet designed by Oak Ridge National Laboratory to evaluate the interaction between UO2 and various FeCrAl alloys revealed no extensive interaction in all the samples irradiated in the rodlet. Localized defects, namely circular pitting and cracks of a few tents of microns, were observed on the FeCrAl disc surfaces in some of the diffusion couples. These defects might be surface fabrication defects rather than irradiation-induced localized corrosion, but chemical analyses are needed to determine whether accumulation of corrosive fission products are causing enhanced interaction. Both non-destructive and destructive PIEs were collected on a rodlet containing U3Si5 fuel. The rodlet had extremely low fission gas release, similar to previously analyzed U3Si2 rodlets. Gamma tomography revealed thermo-migration of fission product ruthenium toward the center of the fuel pellet. The fuel microstructure resembles the as-fabricated microstructure, showing extensive microcracking related to the phase transition at 450°C. Two types of secondary phases were observed, which could either be due to segregation and precipitation of fission products or to UN impurities from fabrication. Finally, the current status of PIE on other ATF-1 irradiations will be discussed. These irradiations include UN-U3Si5 rodlets, UN-U3Si2 rodlets and an additional test with U3Si2 fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Low absorption loss p-AlGaN superlattice cladding layer for current-injection deep ultraviolet laser diodes

Current injection into AlGaN-based laser diode structures with high aluminum mole fractions for deep ultraviolet emission is investigated. The electrical characteristics of laser diode structures with different p-AlGaN short period superlattice (SPSL) cladding layers with various aluminum mole fractions are compared. The heterostructures contain all elements that are needed for a current-injection laser diode including cladding and waveguide layers as well as an AlGaN quantum well active region emitting near 270 nm. We found that with increasing aluminum content in the p-AlGaN cladding, the diode turn-on voltage increases, while the series resistance slightly decreases. By introducing an SPSL instead of bulk layers, the operating voltage is significantly reduced. A gain guided broad area laser diode structure with transparent p-Al{sub 0.70}Ga{sub 0.30}N waveguide layers and a transparent p-cladding with an average aluminum content of 81% was designed for strong confinement of the transverse optical mode and low optical losses. Using an optimized SPSL, this diode could sustain current densities of more than 4.5 kA/cm{sup 2}.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2017

Methodology of Experiment Design for Improving Transient Critical Heat Flux Model Development

One phenomenon that contributes to nuclear fuel and clad failure mechanisms is the departure from nucleate boiling (DNB). Accident scenarios such as reactivity-initiated accidents (RIAs) provide conditions where the frequency of boiling crisis is much more likely to occur and on much shorter time scales than loss of coolant (LOCA) or loss of flow (LOFA) accidents. Here, boiling crisis in the form of DNB is attributed to transient-induced heat flux values that exceed the critical heat flux (CHF). The resulting effects, such as bubble crowding and eventual dryout, can affect flow boiling heat transfer and compromise the structural integrity of nuclear reactor materials. The focus of the work proposed here lies in developing a rigorous process using out-of-pile testing, under simulated RIA conditions, to determine the transient critical heat flux of cladding materials, without contributions from aforementioned radiation-induced effects that have been observed during in-pile experiments. This study provides a road-map for developing and performing experiments that simulate the power profile and energy deposition aspects of transient events, specific to RIAs, on the laboratory scale. The goal of our project is to observe and measure DNB in an annular flow channel around a single heater rod that is operating and nominal thermal hydraulic conditions of a PWR. It is also an aim of these experiments to determine how to quantify the gap between SS and transient CHF cases, based on state-dependent properties of the cladding material type that is being interrogated. Traditionally, CHF is determined as a combination of four factors: hydraulic diameter, system pressure, mass flow, fluid quality. The authors propose that an additional measureable quantity, which is dependent on the structural properties of the cladding material itself, is partly reasonable for the difference between measured and model-developed CHF values. Using a carefully metered approach will allow for several key concepts to be realized: compensation for transient-based effects to current models for CHF; adaptation of current models for different cladding material types, including Gen-IV concept materials; and implementation of a correction factor to account for both power transient-based and radiation-induced effects on the CHF that have not yet been accounted for.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

R&D Plan for RISMC Industry Application #1: ECCS/LOCA Cladding Acceptance Criteria

The Nuclear Regulatory Commission (NRC) is finalizing a rulemaking change that would revise the requirements in 10 CFR 50.46. In the proposed new rulemaking, designated as 10 CFR 50.46c, the NRC proposes a fuel performance-based equivalent cladding reacted (ECR) criterion as a function of cladding hydrogen content before the accident (pre-transient) in order to include the effects of higher burnup on cladding performance as well as to address other technical issues. A loss of operational margin may result due to the more restrictive cladding embrittlement criteria. Initial and future compliance with the rule may significantly increase vendor workload and licensee costs as a spectrum of fuel rod initial burnup states may need to be analyzed to demonstrate compliance. The Idaho National Laboratory (INL) has initiated a project, as part of the DOE Light Water Reactor Sustainability Program (LWRS), to develop analytical capabilities to support the industry in the transition to the new rule. This project is called the Industry Application 1 (IA1) within the Risk-Informed Safety Margin Characterization (RISMC) Pathway of LWRS. The general idea behind the initiative is the development of an Integrated Evaluation Model (IEM). The motivation is to develop a multiphysics framework to analyze how uncertainties are propagated across the stream of physical disciplines and data involved, as well as how risks are evaluated in a LOCA safety analysis as regulated under 10 CFR 50.46c. This IEM is called LOTUS which stands for LOCA Toolkit for US, and it represents the LWRS Program’s response to the proposed new rule making. The focus of this report is to complete an R&D plan to describe the demonstration of the LOCA/ECCS RISMC Industry Application # 1 using the advanced RISMC Toolkit and methodologies. This report includes the description and development plan for a RISMC LOCA tool that fully couples advanced MOOSE tools already in development in order to characterize and optimize plant safety and operational margins. Advanced MOOSE tools that are needed to complete this integrated evaluation model are: RAVEN, RELAP-7, BISON, and MAMMOTH.

10CFR50.46C↗
DOE OSTI2018

Fuel – clad chemical interaction evaluation of the TREAT reactor conceptual low-enriched-uranium fuel element

The Transient Reactor Test (TREAT) facility resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL). The TREAT reactor is currently undergoing design and engineering studies for its conversion from a high enriched uranium (HEU) to a low-enriched uranium (LEU) core. The conceptual design of the LEU fuel element identified two main design differences compared with the HEU fuel element; namely, it will contain four times more fissionable material in its graphite matrix and distinct nuclear-grade Zirconium alloy, as Zircaloy-3 was used in the HEU fuel assembly and is not commercially available currently. These design changes may impact the magnitude of chemical interaction between fuel and cladding materials during physical contact under expected TREAT operation conditions and, therefore, was evaluated through a combination of experimental testing and thermodynamic modeling in order to determine implications for the fuel assembly. In this study, two potential cladding material types, Zircaloy-4 or Zr-1Nb alloys, were evaluated, and it was found for both material types that the extent of interaction and specific chemical reactions are minimal and no detrimental effect on the overall cladding properties is observed. Here, the resulting interaction layer of 3-6 µm was measured after a 2-week exposure at 820°C. The thermodynamic analysis was extended to temperatures beyond the TREAT reactor operation and accident conditions in order to give some insight that may be of interest for other reactor systems as the High Temperature Gas Reactors (operation above 1,000°C) and for Nuclear Reactor Severe Accident phenomenology study where the UO 2 fuel could reach temperatures over 2,800°C and melt.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗