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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 541 records · Page 30
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Analysis of iron-chromium-aluminum samples exposed to accident conditions followed by quench in the QUENCH-19 experiment

The QUENCH-19 experiment was a first-of-its-kind full-bundle test simulating accident conditions followed by water quench on accident-tolerant fuel (ATF) cladding. Here, a type of FeCrAl(Y) alloy, B136Y3, was developed at Oak Ridge National Laboratory and tested at the Karlsruhe Institute of Technology using Kanthal APM corner rods, a shroud, and Kanthal AF spacer grids. Testing conditions were similar to those in QUENCH-15—which tested ZIRLO cladding behavior—so that B136Y3 and ZIRLO cladding could be compared. QUENCH-19 consisted of an initial pre-oxidation heating followed by a transient. Then, a maximum power hold, which was not present in QUENCH-15, was executed to extend the heating period for the FeCrAl(Y) rods. Finally, a rapid water quench was executed that was similar to emergency core coolant system (ECCS) actuation. Compared with the ZIRLO rods in QUENCH-15, the bundle in QUENCH-19 released significantly less H 2 (9.2 g vs. 47.6 g) and achieved a much lower maximum temperature (1455°C vs. 1880°C). Furthermore, no breakaway oxidation was observed in QUENCH-19. Metallographic mounts revealed that despite the symmetry of the setup, at elevations near the maximum temperature, cladding and thermocouples were heavily damaged, substantial melting and oxidation occurred, and the cladding underwent chemical interaction with the thermocouple sheaths. Additionally, the ZrO 2 spacers detrimentally interacted with the cladding, leading to mixed oxide debris and the full destruction of some rods. Additional failure was found in certain cooler rods that may have risen due to the high thermal expansion coefficient of FeCrAl alloys. This paper presents an analysis of this work, which suggests that FeCrAl cladding can chemically survive anticipated loss-of-coolant accident events followed by rapid ECCS quench if the correct geometry and core design are present.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Plutonium-238 Production Program Results, Implications, and Projections from Irradiation and Examination of Initial NpO 2 Test Targets for Improved Production

An alternative target design with potential improvements, including a major increase in 238 Pu production rate and annual capacity; fewer targets to be fabricated, irradiated, and processed; and a significant replacement of a large volume of caustic-nitrate, aluminum-bearing radioactive liquid waste with a smaller volume of solid metal waste, has been conceived and evaluated using reactor physics and thermal-hydraulic analyses. The alternative target design uses pressed pellets of 237 NpO 2 , sintered to 92% to 93% of theoretical density, and stacked inside a Zircaloy-4 cladding tube. Additionally, four test targets were fabricated, irradiated, and examined. No melting or other potential problems were indicated. Projections from measured constituents indicated annual production could be increased by a factor of ~2, and the number of targets required to be fabricated, irradiated, and processed could be reduced by a factor of ~5.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

IER 296 Experiment- Thermal Feasibility Analysis [Slides]

This set of slides summarizes progress, highlights, and issues related to the LANL design engineering effort for the IER 296 critical experiment. The experiment configuration described in Section 5.1.1 of CED-1 IRSN Preliminary Design for Thermal/Epithermal Experiments with MOX FUEL (TEX-MOX) is re-examined in a finite element analysis model using design geometry utilized in the Chlorine Worth Study (CWS) Experiments. Preliminary results and design considerations are presented. Additional analysis has been performed based on comments from the collaborative virtual meeting on 12JAN2023. Specifically, the fuel cladding and gas gap have been included in the thermal analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

In-Situ Powder-Directed Energy Process Control for Additively Manufactured Multi-Layer, Functionally Graded Components

Coatings and cladding can experience damage in harsh environments: Erosion Oxidation Delamination External mechanical damage Radiation induced damage Functionally Graded Materials (FGMs) were primarily designed to reduce thermal stress in thermal barrier coatings (TBCs), but can produce coatings which resist a variety of material damage (corrosion, surface damage, diffusion) Powder-fed DED with multiple independent feeders allows for AM FGMs, and provides robust control over process parameters to optimize microstructure and performance. FGMs have a great opportunity for applications in energy applications as a strategy to increase plant efficiency and lifetime.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

An Innovative Accident Tolerant LWR Fuel Rod Design Based on Uranium-Molybdenum Metal Alloy

The US Department of Energy is developing a uranium-molybdenum metal alloy Enhanced Accident Tolerant Fuel concept for Light Water Reactor applications that provides improved fuel performance during normal operation, anticipated operational occurrences, and postulated accidents. The high initial uranium atom density, the high thermal conductivity, and a low heat capacity permit a U-Mo-based fuel assembly to meet important design and safety requirements. These attributes also result in a fuel design that can satisfy increased fuel utilization demands and allow for improved accident tolerance in LWRs. This paper summarizes the results obtained from the on-going activities to; 1) evaluate the impact of the U-10wt%Mo thermal properties on operational and accident safety margins, 2) produce a triple extrusion of stainless steel cladding/niobium liner/U-10Mo fuel rod specimen and 3) test the high temperature water corrosion of rodlet samples containing a drilled hole in the cladding. Characterization of the cladding and liner thickness uniformity, microstructural features of the U-Mo gamma phase, and the metallurgical bond between the component materials will be presented. The results from corrosion testing will be discussed which yield insights into the resistance to attack by water ingress during high temperature water exposure for the triple extruded samples containing a drilled hole. These preliminary evaluations find that the U-10Mo fuel design concept has many beneficial features that can meet or improve conventional LWR fuel performance requirements under normal operation, AOOs, and postulated accidents. The viability of a deployable U-Mo fuel design hinges on demonstrating that fabrication processes and alloying additions can produce acceptable irradiation stability during normal operation and accident conditions and controlled metal-water reaction rates in the unlikely event of a cladding perforation. In the area of enhanced accident tolerance, a key objective is to establish that the lower stored energy of the U-Mo fuel design can provide the emergency core cooling systems the opportunity to maintain the reactor core in a coolable geometry following an accident.

Accident tolerance fuel↗
DOE OSTI2015

Report On Design And Preliminary Data Of Halden In-Pile Creep Rig

A set of in-pile creep tests is ongoing in the Halden reactor on ORNL’s candidate accident tolerant fuel cladding materials. These tests are meant to provide essential material property information that is needed for an informed analysis of these fuel concepts under normal operating conditions. These tests provide detailed information regarding swelling, thermal creep, and irradiation creep rates of these materials. The results to date have been compared with the limited set of information available in literature that is form irradiation tests in other reactors or out-of-pile tests. Most of the results are in good agreement with prior literature, except for irradiation creep rate of SiC. To elucidate the difference between the HFIR and Halden test results continued testing is necessary. The tests describe in this progress report are ongoing and will continue for at least another year.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Aluminum Clad Spent Nuclear Fuel Task 6: Surrogate Sample Preparation and Validation Experiment Test Plan

This document presents the experimental test plan for Task 6: Surrogate Preparation and Validation. Task 6 will produce and validate surrogate oxyhydroxide materials used in other experimental tasks within the larger Aluminum Clad Spent Nuclear Fuel Long Term Dry Storage Project. Task 6 includes both the preparation and characterization of surrogate oxyhydroxide materials as well as the analysis of oxide/oxyhydroxide layers on Savannah River Site Spent Nuclear Fuel Materials. This test plan provides experimental design details and the task’s overall quality assurance approach.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

In-line Mach-Zehnder interferometer composed of microtaper and long-period grating in all-solid photonic bandgap fiber

We report a compact in-line Mach-Zehnder interferometer combining a microtaper with a long-period grating (LPG) in a section of all-solid photonic bandgap fiber. Theoretical and experimental investigations reveal that the interferometer works from the interference between the fundamental core mode and the LP{sub 01} cladding supermodes. The mechanism underlying the mode coupling caused by the microtaper can be attributed to a bandgap-shifting as the fiber diameter is abruptly scaled down. In addition, the interferometer designed to strengthen the coupling ratio of the long-period grating has a promising practical application in the simultaneous measurement of curvature and temperature.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2016

Advanced Characterization of Fuel Pellet and Cladding from the HB Robinson NPP

In order to predict the behavior and response of materials to extreme environments of a nuclear reactor, comprehensive models have been developed over the past several decades. These models are typically validated against carefully designed single or double variable experiments which may be systematically altered in order build up a model of behavior under the conditions being examined. In reality, this does not necessarily compare well with the real application situations whereby the behavior of the material under irradiation is determined by multiple variables changing simultaneously e.g. temperature, irradiation flux and fluence, and pressure. This is particularly so when considering the highly dynamic environment of a fuel assembly within a nuclear power plant (NPP) core. This work aims to provide near-atomic resolution microstructural and chemical data of a fuel/clad specimen that has been used in service within a commercial NPP. Atom probe tomography and S/TEM characterization of the cladding showed the presence of small scale clusters containing Fe and Cr, along with segregation of these minor solutes to dislocations. SEM examination of the fuel pellet showed relatively large fission gas bubbles that had burst due to the metallographic polishing process to prepare the sample for examination. SEM-EDX mapping revealed the presence of Xr bubbles located beneath the surface that still retained their Xe concentration. Metallic precipitates were also observed. S/TEM-EDX of the fuel material also revealed the presence of small metallic precipitates. In addition, small Xe bubbles were also found to be present in the material, lying along grain boundaries only. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Development of Instrumented Advanced Test Reactor Irradiation Capsule Experiment for In-situ Thermal Conductivity Measurements of High-Density Fuels

Idaho National Laboratory (INL) is developing a first-of-a-kind leadout instrumented capsule experiment design to enable in-situ measurement capabilities in the Advanced Test Reactor (ATR) core. The Ceramic Advanced Thermal Evolution Research (CRATER) experiment supports the aLEU program objective to accelerate fuel performance irradiation testing for identifying alternative high-assay, low enriched uranium (HALEU) fuel systems. CRATER is a fueled, instrumented capsule experiment to measure in-situ temperature and thermal conductivity of ceramic fuels. Two ceramic fuel types will be used, uranium mono-nitride (UN) and uranium mono-carbide (UC), with a third metallic fuel used for comparison (UMo). The three fuel specimens will use a stainless-steel cladding. Programmatic objectives include linear heat generation rates (LHGR) of 210 ± 25 Watts per cm. and an inner clad temperature of 300-450 °C. The evolution of fuel thermal conductivity during irradiation has never been successfully measured in-situ for these systems and this experiment is designed to use advances in measurement sciences to characterize how thermal transport properties evolve while in reactor. Neutronic simulations of the experiment and its surrounding reactor environment were conducted using the Monte Carlo N-Particle Transport code (MCNP) and result in optimized fuel enrichment to meet target linear heat generation rates (LHGRs) influencing fuel temperatures, and fuel burnup requirements. Fabrication research and development (R&D) efforts are underway to produce annular right cylinder UC and UN pellets using carbothermic reduction and nitridation (or hydride-dehydride-nitride) synthesis methods, followed double-action die cold isostatic pressing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Demonstration of LOTUS multiphysics BEPU analysis framework for LB-LOCA simulations

The LOCA Toolkit for U.S. light water reactors (LOTUS) currently under development at Idaho National Labs (INL) is a plug and play, multiphysics environment to be used in support of system level plant transient analysis. New proposed rule changes in LOCA safety regulations (10 CFR 50.46c) require the inclusion of cladding hydrogen content in the evaluation of equivalent cladding reacted (ECR) and peak cladding temperature (PCT). The potential for a future implementation of said rule change motivates the current development of a unique union of core design, system analysis, and fuel performance codes within a framework allowing for uncertainty quantification (UQ). A demonstration of LOTUS capabilities to address this potential need is performed with an integration of the core design code PHISICS, the fuel performance code FRAPCON, and the system analysis code RELAP5-3D. UQ is performed via Wilks’ method to compute the one sided 95/95 confidence values of the aforementioned safety metrics. Results demonstrate the benefits of the Multiphysics Best Estimate Plus Uncertainty (MP-BEPU) methodology and provide useful visualization of the limiting cases.

10CFR50.46c↗
DOE OSTI2023

A fine pore-preserved deep neural network for porosity analytics of a high burnup U-10Zr metallic fuel

Abstract U-10 wt.% Zr (U-10Zr) metallic fuel is the leading candidate for next-generation sodium-cooled fast reactors. Porosity is one of the most important factors that impacts the performance of U-10Zr metallic fuel. The pores generated by the fission gas accumulation can lead to changes in thermal conductivity, fuel swelling, Fuel-Cladding Chemical Interaction (FCCI) and Fuel-Cladding Mechanical Interaction (FCMI). Therefore, it is crucial to accurately segment and analyze porosity to understand the U-10Zr fuel system to design future fast reactors. To address the above issues, we introduce a workflow to process and analyze multi-source Scanning Electron Microscope (SEM) image data. Moreover, an encoder-decoder-based, deep fully convolutional network is proposed to segment pores accurately by integrating the residual unit and the densely-connected units. Two SEM 250 × field of view image datasets with different formats are utilized to evaluate the new proposed model’s performance. Sufficient comparison results demonstrate that our method quantitatively outperforms two popular deep fully convolutional networks. Furthermore, we conducted experiments on the third SEM 2500 × field of view image dataset, and the transfer learning results show the potential capability to transfer the knowledge from low-magnification images to high-magnification images. Finally, we use a pre-trained network to predict the pores of SEM images in the whole cross-sectional image and obtain quantitative porosity analysis. Our findings will guide the SEM microscopy data collection efficiently, provide a mechanistic understanding of the U-10Zr fuel system and bridge the gap between advanced characterization to fuel system design.

36 MATERIALS SCIENCE↗
DOE OSTI2019

First-principles calculations of solute transport in zirconium: Vacancy-mediated diffusion with metastable states and interstitial diffusion

Zirconium alloys are the most widely used nuclear fuel cladding materials for light water power reactors where irradiation damage causes solute redistribution, leading to degradation of alloy properties such as corrosion resistance. Designing radiation-tolerant zirconium alloys requires a thorough understanding of the atomic-scale transport behavior of the alloying elements in Zr. In this work, we perform density function theory calculations to investigate the diffusion of Sn, Cr, Fe, Be, Al, and Ni in the hexagonal close-packed (HCP) Zr matrix. We develop a methodology to accurately model the metastable vacancy states along the basal migration path, known to occur in group IV metals. We compute the vacancy-mediated solute diffusion coefficients and drag ratios using the kinetic Monte Carlo method and an analytic Green's function method—the agreement between the two validates our methodology. The computed diffusion coefficients of Sn and Al show good agreement with the experimental data and we expect these solutes to diffuse via the vacancy-mediated mechanism. We use a Green's function approach, parameterized with data from density functional theory calculations, to compute the interstitial diffusion coefficients of Cr, Fe, Be, and Ni in the HCP Zr lattice. The computed diffusion coefficients of Cr, Ni, and Be agree with the experimental measurements within one order of magnitude, while those of Fe are within two orders of magnitude of the experimental measurements. The drag ratios for Cr, Fe, Be, and Ni are positive up to 1235 K, which suggests that nonequilibrium vacancy fluxes could drag these solutes toward sinks such as dislocation loops and grain boundaries. We also compute the transport coefficients without including the metastable states, and using the eight- and thirteen-frequency model. Our results show significant differences in drag ratio for the eight- and thirteen-frequency model predictions compared with the Green's function methodology, but smaller errors in the solute diffusivity. Combining interstitial and vacancy-mediated diffusivities, we predict the unusual result that increased vacancy concentration slows down solute diffusivity, while a sufficiently high vacancy concentration can change the dominant mechanism to an accelerated vacancy-mediated diffusion.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Milestone 1.2.13: Preliminary Measurements of Radiolytic Nitric Acid Formation to Support Predictive Model Validation

Predictive computational models have been developed to support the technical basis for extended dry storage of aluminum-clad spent nuclear fuel (ASNF) in helium-backfilled cannisters. To date, these models have been optimized on a variety of irradiation experiments designed to elucidate the radiation-induced formation of molecular hydrogen gas, a radiolysis product that is potentially problematic for the safe storage of ASNF. However, the yield of nitric acid (HNO3) has not been determined, despite conservative predictions of its formation (300–4000 ppm) in 1% residual air environments irradiated in contact with ASNF. HNO3, another problematic radiolysis product, can lead to enhanced corrosion and potentially compromise storage canister integrity. Thus, to support the validation of predictive computer models, we report the measurement and quantification of HNO3 from the gamma irradiation of aluminum alloy coupons in humid air.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2016

Role of Defects in Swelling and Creep of Irradiated SiC

Silicon carbide is a promising cladding material because of its high strength and relatively good corrosion resistance. However, SiC is brittle and therefore SiC-based components need to be carefully designed to avoid cracking and failure by fracture. In design of SiC-based composites for nuclear reactor applications it is essential to take into account how mechanical properties are affected by radiation and temperature, or in other words, what strains and stresses develop in this material due to environmental conditions. While thermal strains in SiC can be predicted using classical theories, radiation-induced strains are much less understood. In particular, it is critical to correctly account for radiation swelling and radiation creep, which contribute significantly to dimensional instability of SiC under radiation. Swelling typically increases logarithmically with radiation dose and saturates at relatively low doses (damage levels of a few dpa). Consequently, swelling-induced stresses are likely to develop within a few months of operation of a reactor. Radiation-induced volume swelling in SiC can be as high as 2%, which is significantly higher than the cracking strain of 0.1% in SiC. Swelling-induced strains will lead to enormous stresses and fracture, unless these stresses can be relaxed via some other mechanism. An effective way to achieve stress relaxation is via radiation creep. Although it has been hypothesized that both radiation swelling and radiation creep are driven by formation of defect clusters, existing models for swelling and creep in SiC are limited by the lack of understanding of specific defects that form due to radiation in the range of temperatures relevant to fuel cladding in light water reactors (LWRs) (<1000°C). For example, defects that can be detected with traditional transmission electron microscopy (TEM) techniques account only for 10-45% of the swelling measured in irradiated SiC. Here, we have undertaken an integrated experimental and modeling effort to discover the previously invisible defects in irradiated SiC and to determine the contributions of these defects to radiation swelling. Knowledge of the most stable defect structures and the rate controlling processes during defect evolution is essential for development of predictive models for swelling and creep as a function of temperature and radiation dose. This research has been enabled by state-of-the-art imaging techniques, such as the aberration corrected scanning transmission electron microscopy (STEM) (FEI TITAN) closely coupled with multi-scale models of stable defect clusters and their evolution.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Comparison of Two High-Performance Energy Efficient Homes: Annual Performance Report, December 1, 2010–November 30, 2011

Beginning in 2008, two pairs of energy-saver houses were built at Wolf Creek in Oak Ridge, TN. These houses were designed to maximize energy efficiency using new ultra-high-efficiency components emerging from ORNL s Cooperative Research and Development Agreement (CRADA) partners and others. The first two houses contained 3713 square feet of conditioned area and were designated as WC1 and WC2; the second pair consisted of 2721 square feet conditioned area with crawlspace foundation and they re called WC3 and WC4. This report is focused on the annual energy performance of WC3 and WC4, and how they compare against a previously benchmarked maximum energy efficient house of a similar footprint. WC3 and WC4 are both about 55-60% more efficient than traditional new construction. Each house showcases a different envelope system: WC3 is built with advanced framing featured cellulose insulation partially mixed with phase change materials (PCM); and WC4 house has cladding composed of an exterior insulation and finish system (EIFS). The previously benchmarked house was one of three built at the Campbell Creek subdivision in Knoxville, TN. This house (CC3) was designed as a transformation of a builder house (CC1) with the most advanced energy-efficiency features, including solar electricity and hot water, which market conditions are likely to permit within the 2012 2015 period. The builder house itself was representative of a standard, IECC 2006 code-certified, all-electric house built by the builder to sell around 2005 2008.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2017

Effects of fuel thickness on thermal and mechanical performance of U10-Mo monolithic fuel plates

Monolithic fuels are being considered for conversion of high performance research and test reactors. These plate-type fuels are comprised of a low enrichment, high density U-Mo alloy fuels within an aluminum cladding. Although the plates have demonstrated satisfactory performance; still, use of a high density fuel in a foil form retains technical challenges that could affect the overall performance. To understand performance of this new design during fabrication and irradiation, the plates have been evaluated for various geometrical and operational variables. As a part of these set of parametric sensitivity studies, effects of foil thickness on performance were studied. Because high performance research reactors will utilize different fuel thicknesses, it is necessary to evaluate possible effects for a range of thicknesses that are being considered for various designs. Based on the preliminary design specifications, the fuel thicknesses were varied between the limiting cases. The bounding fuel thicknesses were 0.203 mm as minimum that is projected for ATR, and 0.635 mm as maximum that is projected for MITR. To study possible effects, a behavioral model was developed for a selected plate from RERTR-12 experiments and the plate was simulated with as run irradiation history. The simulations were repeated for a range of fuel thicknesses, while keeping the cladding thickness and the operational parameters the same. The results have indicated that the plates with thicker fuels would have higher temperatures, deformations and shutdown stresses. To investigate the effects of fuel thickness exclusively, operational parameters were scaled. In particular, a second set of simulations with prorated volumetric heat generation rates were performed. The results indicated that the fabrication stresses in the fuel foil decrease with an increasing fuel thickness. On contrary, irradiation stresses of the fuel at shutdown are higher for the plates with thicker fuels. Greater peak deformations occurred in plates with thicker fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗