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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 613 records · Page 34
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Investigation of δ zirconium hydride morphology in a single crystal using quantitative phase field simulations supported by experiments

In light water nuclear reactors, waterside corrosion of the cladding material leads to the production of hydrogen, a fraction of which is picked up by the zirconium cladding and precipitates into brittle hydride particles. These nanoscale hydride particles aggregate into mesoscale hydride clusters. The principal stacking direction of the nanoscale hydrides precipitated in the cladding tube changes from circumferential in the absence of applied stress to radial under circumferential applied stress. A quantitative phase field model has been developed to predict the hydride morphology observed experimentally and identify the mechanisms responsible for nanoscale hydride stacking. The model focuses on nanoscale hydride precipitation in a single zirconium grain with a detailed description of the anisotropic elastic contribution. The model predictions concerning the shape, orientation, and stacking behavior of nanoscale hydride are analyzed and compared with experimental observations. The model accurately accounts for the experimentally observed elongated nanoscale hydride shape and the stacking of hydrides along the basal plane of the hexagonal zirconium matrix. When investigating the role of applied stress in hydride morphology, the model challenges some of the mechanisms previously proposed to explain hydride reorientation. Finally, although hydride reorientation has been hypothesized to be caused by a change in nanoscale hydride shape, the current study shows that these mechanisms are unlikely to occur.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Development of Cr-coated zirconium alloy cladding for enhanced accident tolerance

In the context of the development of Enhanced Accident Tolerant Fuels (EATF), AREVA NP, along with CEA and EDF, is developing, among others, a chromium (Cr) coating (5- 20 μm thick) on zirconium alloy cladding. The coating is deposited on the cladding surface by Physical Vapor Deposition (PVD) and forms a dense and adherent protective layer. It preserves the good mechanical behavior of the current cladding in normal operating conditions while significantly improving the oxidation behavior in accidental conditions. This behavior is being evaluated through out-of-pile corrosion and mechanical tests at the AREVA NP Technical and Research Centers. The corrosion behavior was evaluated through autoclave tests in both PWR water chemistry at 360 and in 415 Celsius degrees steam to increase the corrosion kinetics. In both cases very small weight gain was measured on the order of 1 mg/dm{sup 2} and remains stable throughout the length of the tests. The mechanical behavior of Cr-coated samples was evaluated through tensile tests on Cr-coated and uncoated M5 tube samples at both room temperature and 400 Celsius degrees. The curves demonstrate that the Cr-coated samples exhibit similar mechanical behavior as uncoated samples, and for all the tests performed, the mechanical properties (yield strength 0.2%, ultimate tensile strength, and total elongation) of the Cr-coated samples all fell within the range of values for uncoated samples. In order to obtain in-pile data as early as possible on this solution, an irradiation program (called IMAGO) in the Goesgen reactor will be performed in the years to come. This will be the first irradiation of EATF solutions in a commercial reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Report Summarizing the Mechanical Properties of a Large ODS Ferritic Alloy Ingot by Forging at High Temperatures for Future Mother Tube Production

Oxide dispersion strengthened (ODS) ferritic alloys are considered the benchmark fuel cladding and core structural material in advanced nuclear energy reactors that require high-temperature strength and creep properties and resistance to radiation damage. The ODS ferritic alloys including 14YWT are produced by mechanical alloying (MA), which is time-consuming and is associated with high manufacturing costs that are not beneficial to being used in advanced nuclear energy reactors. This report summarizes the accomplishments in FY24 in the INM program for producing an ODS ferritic alloy by high-deformation, high-temperature processing of reactive and ferritic alloy powders for advanced reactor fuel cladding applications. Following four hot forging experiments, it was concluded that several significant challenges were encountered that were difficult to overcome. These challenges were related to using high annealing temperatures for pressure assisted sintering processes to produce dense microstructures, but the high annealing temperatures result in long range diffusion of Ti through the bcc Fe lattice and react with the YIG particles that are distributed on the prior surfaces of the ferritic alloy powders. In addition, it was determined that the high deformations induced by forging at high temperatures were not very effective for fracturing the reactive YIG particles into smaller particles and distributing them into the interior of the ferritic alloy powders. Spark plasma sintering was attempted for shortening the time at high temperatures, which led to full densification but the microstructure characterization results still showed that Ti atoms diffused over long distances in the bcc Fe lattice and reacted with the YIG particles. Finally, a short 30 minute high intensity ball milling experiment was conducted on blended 14WT and YIG powders for inducing severe deformations that caused the initially spherical powders of 14WT decorated with the YIG particles on the surfaces to transform to flakes and fracturing of YIG particles into smaller particles that were incorporated into the bcc Fe lattice. Two forgings were performed with the high intensity milled powders: first annealing for 20 minutes at 850ºC followed by forging and second annealing for 20 minutes at 1,100ºC followed by forging. At 850ºC, Ti atoms are effectively immobile, thus allowing for densification albeit incomplete. The deformation by forging after annealing for 20 minutes at 1,100ºC favored dynamic recrystallization processes that led to nano-size grains with very high stored energy due to high dislocation density. The corresponding VH data showed significant hardening that correlated with estimated strengthening of ~1650 MPa. This hybrid processing approach combining high intensity ball milling of powder with annealing and forging showed the most promise for producing an oxide dispersing strengthened ferritic alloy.

36 MATERIALS SCIENCE↗
DOE OSTI2016

SiC cladding corrosion and mitigation

Silicon carbide (SiC) is a leading candidate for accident tolerant fuel. Testing of SiC in autoclaves at 360 C. degrees in simulated coolant resulted in little corrosion. However, testing of SiC tubes in the Massachusetts Institute of Technology (MIT) reactor manufactured in the same way had unacceptable corrosion levels. This in-reactor SiC corrosion issue is now being addressed by Westinghouse in tests being carried out at our Churchill, PA facility and in tests in the MIT reactor. The suspected driver for the excessive corrosion is the generation of oxidative water radiolysis products in the core. The MIT reactor was identified as running at H{sub 2} levels of ∼20 to 22 cc/kg coolant (cubic centimeters standard temperature and pressure per kilogram coolant) which were below those normally specified in operating pressurized water reactors (PWRs) of 25 to 50 cc/kg. Hydrogen is used to react with the radiolysis products to reduce them to water. The current MIT reactor tests will be run at H{sub 2} concentration levels of ∼50 cc/kg. The applicability of this approach has been supported by recent papers in the literature. Parallel testing at Churchill is being carried out to verify the efficacy of this approach using autoclaves operating at 345 C.degrees with H{sub 2}O{sub 2} injection to simulate the water radiolysis products. Various concentrations of H{sub 2} are being tested to determine the effect on SiC corrosion rate. Further studies are underway to investigate the effects of different SiC fabrication and structural parameters and to optimize for improved corrosion performance. This paper compares the results of the corrosion tests in the MIT reactor and the autoclave tests with H{sub 2}O{sub 2} injection and the effect of H{sub 2} injection on the SiC corrosion rate. The results of the autoclave tests with H{sub 2}O{sub 2} addition indicate that corrosion rates for SiC can be achieved that are much higher than those achieved in a reactor. It appears that the use of H{sub 2}O{sub 2} addition in flow through loops to simulate the corrosion rate of SiC samples in reactor is feasible. The use of H{sub 2} addition to reduce the corrosion rate in-reactor of SiC appears to be feasible.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Assembly and Delivery of Rabbit Capsules for Irradiation of Silicon Carbide Cladding Tube Specimens in the High Flux Isotope Reactor

Neutron irradiation of silicon carbide (SiC)-based fuel cladding under a high radial heat flux presents a critical challenge for SiC cladding concepts in light water reactors (LWRs). Fission heating in the fuel provides a high heat flux through the cladding, which, combined with the degraded thermal conductivity of SiC under irradiation, results in a large temperature gradient through the thickness of the cladding. The strong temperature dependence of swelling in SiC creates a complex stress profile in SiCbased cladding tubes as a result of differential swelling. The Nuclear Science User Facilities (NSUF) Program within the US Department of Energy Office of Nuclear Energy is supporting research efforts to improve the scientific understanding of the effects of irradiation on SiC cladding tubes. Ultimately, the results of this project will provide experimental validation of multi-physics models for SiC-based fuel cladding during LWR operation. The first objective of this project is to irradiate tube specimens using a previously developed design that allows for irradiation testing of miniature SiC tube specimens subjected to a high radial heat flux. The previous “rabbit” capsule design uses the gamma heating in the core of the High Flux Isotope Reactor (HFIR) to drive a high heat flux through the cladding tube specimens. A compressible aluminum foil allows for a constant thermal contact conductance between the cladding tubes and the rabbit housing despite swelling of the SiC tubes. To allow separation of the effects of irradiation from those due to differential swelling under a high heat flux, a new design was developed under the NSUF program. This design allows for irradiation of similar SiC cladding tube specimens without a high radial heat flux. This report briefly describes the irradiation experiment design concepts, summarizes the irradiation test matrix, and reports on the successful delivery of six rabbit capsules to the HFIR. Rabbits of both low and high heat flux configurations have been assembled, welded, evaluated, and delivered to the HFIR along with a complete quality assurance fabrication package. These rabbits contain a wide variety of specimens including monolith tubes, SiC fiber SiC matrix (SiC/SiC) composites, duplex specimens (inner composite, outer monolith), and specimens with a variety of metallic or ceramic coatings on the outer surface. The rabbits are targeted for insertion during HFIR cycle 475, which is scheduled for September 2017.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Electron microscopy study of the formation mechanism of catalytic nickel-rich particles and the role of carbonyl sulphide in the suppression of carbon deposition on 20Cr-25Ni steel

Highlights: • Carbon deposition has been significantly reduced under the presence of CoS. • The origin of catalytic nickel-rich particles in stainless steel revealed • High resolution EDS study showed that sulphur was observed to be surrounding nickel-rich particles. - Abstract: Austenitic stainless steel is used as fuel cladding in advanced gas-cooled nuclear reactors (AGR). At elevated temperatures, when the steel is exposed to CO{sub 2} based environments filamentary carbon deposits form on the surface of the steel. This filamentary carbon deposition is known to be catalysed by metallic nickel-rich particles. Adding trace amount of carbonyl sulphide (COS) into the gas mixtures suppresses the carbon deposition. In this current work, it has been shown that at 600 °C, the formation of filamentary carbon was suppressed by the addition of 215 ppb COS to a depositing gas mixture (containing approximately 1000 vppm C{sub 2}H{sub 4}/1% CO/bal. CO{sub 2}) which was known to provide the environment suitable for carbon deposition. Samples exposed to the gas mixtures with and without 215 ppb COS were characterised using electron microscopy techniques to understand the formation mechanism of the nickel-rich particles and the inhibition mechanism due to the addition of COS. Electron diffraction study shows that the nickel-rich particles in the oxide layers assume the same crystallography as that of the austenitic metal underneath, regardless of the COS addition. The current observations also show that the metal-oxide interfaces was nickel-rich and a simple model has been proposed to explain the formation of nickel-rich particles within the subsurface oxide. Furthermore, it was found that when COS was added the surface of the nickel-rich particles in the oxide layer was found to be sulphur-rich by energy dispersive spectroscopy (EDS) on a scanning transmission electron microscope (STEM). It is believed that the surface sulphur adsorption onto the nickel-rich particles, rather than bulk sulphide formation, resulted in the inhibition of carbon deformation on the steel.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Characterization of Hydraulic and Ignition Phenomena of Pressurized Water Reactor Fuel Assemblies: Phase II Test Plan

This report summarizes the strategy and preparations for the second phase in the Sandia Fuel Project (SFP) test program. During this phase, five full-length, prototypic 17×17 PWR fuel assembly will simulate a severe loss-of-coolant-accident in the spent fuel pool whereby the fuel is completely uncovered and heats up until ignition of the cladding occurs. Electrically resistive heaters with Zircaloy cladding will substitute for the spent nuclear fuel in the center assembly. This heated assembly will be placed in the center cell of a 3×3 pool rack and will be surrounded by four unheated mock fuel bundles. This arrangement will imitate the situation of a recently offloaded assembly surrounded by much older and thus lower decay heat assemblies. The designs and plans detailed in this report are based on previous testing efforts and represent the current knowledge base. However, these results are subject to change with new information.

Durbin, Samuel G.↗
DOE OSTI2013

Evaluation of Two CEDA Weatherization Pilot Implementations of an Exterior Insulation and Over-Clad Retrofit Strategy for Residential Masonry Buildings in Chicago

This project examines the implementation of an exterior insulation and over-clad strategy for brick masonry buildings in Chicago. The strategy was implemented at a free-standing two story two-family dwelling and a larger free-standing multifamily building. The test homes selected for this research represent predominant housing types for the Chicago area. High heating energy use typical in these buildings threaten housing affordability. Uninsulated mass masonry wall assemblies also have a strongly detrimental impact on comfort. Significant changes to the performance of masonry wall assemblies is generally beyond the reach of typical weatherization (Wx) program resources. The Community and Economic Development Association of Cook County, Inc. (CEDA) has secured a Sustainable Energy Resources for Consumers (SERC) innovation grant sponsored by the United States Department of Energy (DOE). This grant provides CEDA the opportunity to pursue a pilot implementation of innovative approaches to retrofit in masonry wall enclosures. The exterior insulation and over-clad strategy implemented through this project was designed to allow implementation by contractors active in CEDA weatherization programs and using materials and methods familiar to these contractors. The retrofit measures are evaluated in terms of feasibility, cost and performance. Through observations of the strategies implemented, the research described in this report identifies measures critical to performance as well as conditions for wider adoption. The research also identifies common factors that must be considered in determining whether the exterior insulation and over-clad strategy is appropriate for the building.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2017

EPRI/DOE High-Burnup Fuel Sister Rod Test Plan Simplification and Visualization

The EPRI/DOE High-Burnup Confirmatory Data Project (herein called the “Demo”) is a multi-year, multi-entity test with the purpose of providing quantitative and qualitative data to show if high-burnup fuel mechanical properties change in dry storage over a ten-year period. The Demo involves obtaining 32 assemblies of high-burnup PWR fuel of common cladding alloys from the North Anna Nuclear Power Plant, loading them in an NRC-licensed TN-32B cask, drying them according to standard plant procedures, and then storing them on the North Anna dry storage pad for ten years. After the ten-year storage time, the cask will be opened and the mechanical properties of the rods will be tested and analyzed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

A Review of Advanced Test Reactor Fuel and Assessment of Its Compatibility with the ZIRCEX Chlorination Process

Advanced Test Reactor (ATR) fuel has been identified as a resource for high-assay low-enriched uranium (HALEU) production. A survey was performed on the published literature describing ATR fuel. The geometry of the fuel is complex; different parts of the fuel compact experience differing neutron flux and burnup. The literature is sparse, and access is controlled. Therefore, fundamental studies of fuel reprocessing must use a model fuel that represents the main chemical and structural features. Advanced chlorination, or chlorination with sulfur-chlorine bearing reagents is being investigated as way to separate the fuel from metal matrix alloys. A UAl x alloy will be fabricated with x = 3, 4, and 5. The potential chlorination of individual UAl x intermetallics will be assessed in the advanced chlorination process of Al-8001 and Al-6061 as well as a representative mixture. Initial studies will track the alloying elements of the Al, which are Si, Fe, Cu, Mn, Mg, Cr, Zn, and Ti, in addition to the U itself. Further studies will include fission product simulants. Because advanced chlorination solvents include sulfur, the chemistry of sulfur with major and minor constituents will also be investigated. The experimental work accompanied by neutronic calculations will allow the assessment of the feasibility of advanced chlorination to separate aluminum from uranium. If bench-scale testing appears promising, then small-scale tests in shielded facilities with irradiated cladding, lightly irradiated fuel, and spent nuclear fuel are recommended to track the complete inventory of fissile actinides, fission product impurities, and reagent solids and liquids.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Failure behavior of SiC/SiC composite tubes under strain rates similar to the pellet-cladding mechanical interaction phase of reactivity-initiated accidents

The mechanical response of a nuclear-grade silicon carbide fiber-reinforced silicon carbide matrix (SiC/SiC) composite was investigated under mechanical loading conditions mimicking the pellet-cladding mechanical interaction (PCMI) phase of a reactivity-initiated accident (RIA). In a RIA, cladding deformation and failure can be induced by the rapid thermal expansion of the nuclear fuel. A pulse-controlled modified-burst test was used to investigate RIA-like PCMI scenarios on SiC/SiC composite samples at pulse widths from 12 to 100 ms.The strain-driven nature of the cladding sample deformation was due to the rapid internal pressurization and subsequent expansion of a secondary tube. A digital-image correlation technique was used to measure strains from the speckle-painted outer surface of the tubes. The failure strains of samples tested at slower rates, such as RIA event durations of 52 and 100 ms, showed good agreement with the literature-reported values for similar composites tested at slow strain rates. Additionally, the failure strain showed good agreement with reference expansion-due-to-compression tests at slow strain rate. However, a decrease in the failure strain was determined for the fast-rate (12 ms) tests. This indicated that the failure strain of these composites might be influenced by the strain rate during RIA-like events. In conclusion, the failure strains observed in the tests corresponded to local energy depositions of approximately 50 cal/g UO 2 from hot zero power, with an initial condition of pellet–cladding gap closure prior to the event. In-pile transient testing of these concepts that would result in hoop strain due to PCMI in the range of 0.5–1.0% is recommended.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Developing Tomorrows Nuclear Professionals - Apprentice and Graduate Development at PFCS, Sellafield - 19013

The Pile Fuel Cladding Silo (PFCS) at Sellafield, originally built in 1949/50 to receive and store fuel cladding from reprocessing Windscale Pile fuel, has been used to store this material and other waste from the UK nuclear programme ever since. However, the Silo is unsuitable for indefinite storage of the waste; its age, condition and inventory, combined with significant uncertainty in all these factors, rank the facility among the most hazardous in Western Europe. In 2012, Sellafield Ltd began a highly collaborative relationship with its supply chain partners Bechtel Cavendish Nuclear Solutions (BCNS), to undertake a suite of projects to create a capability for retrieving waste. In doing so, the team has discovered a shared passion for graduate student and apprenticed trades development, which has led to the team goal of becoming the number one destination for training secondments within the associated businesses. This paper describes how the team has leveraged the unique challenge of the project, the extensive experience of the delivery team, and the capabilities of the contributing organisations to create a structure secondment experience aimed at maximizing learning and generating enthusiasm for nuclear clean-up. The scheme was developed and implemented, based on some very simple but consistent feedback from the trainee community on what differentiates a good and bad secondment. The paper includes three case studies from individuals who have participated in the programme and the analysis of overall feedback from all programme trainees. Universally, the scheme has been well received with a significant number of trainees coming back to the project to begin their professional careers. The simple differentiators between a good and bad secondment provide a good basis for those wishing to develop similar schemes. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing

New materials are being considered for fuel, cladding and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine and return irradiated samples for each measurement make this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated under pressurized water reactor coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory.

99 GENERAL AND MISCELLANEOUS↗