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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 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Analysis and Experimental Qualification of an Irradiation Capsule Design for Testing Pressurized Water Reactor Fuel Cladding in the High Flux Isotope Reactor

The Advanced Fuels Campaign within the Fuel Cycle Research and Development program of the Department of Energy Office of Nuclear Energy is currently investigating a number of advanced nuclear fuel cladding concepts to improve the accident tolerance of light water reactors. Alumina-forming ferritic alloys (e.g., FeCrAl) are some of the leading candidates to replace traditional zirconium alloys due to their superior oxidation resistance, provided no prohibitive irradiation-induced embrittlement occurs. Oak Ridge National Laboratory has developed experimental designs to irradiate thin-walled cladding tubes with representative pressurized water reactor geometry in the High Flux Isotope Reactor (HFIR) under relevant temperatures. These designs allow for post-irradiation examination (PIE) of cladding that closely resembles expected commercially viable geometries and microstructures. The experiments were designed using relatively inexpensive rabbit capsules for the irradiation vehicle. The simplistic designs combined with the extremely high neutron flux in the HFIR allow for rapid testing of a large test matrix, thus reducing the time and cost needed to advanced cladding materials closer to commercialization. The designs are flexible in that they allow for testing FeCrAl alloys, stainless steels, Inconel alloys, and zirconium alloys (as a reference material) both with and without hydrides. This will allow a direct comparison of the irradiation performance of advanced cladding materials with traditional zirconium alloys. PIE will include studies of dimensional change, microstructure variation, mechanical performance, etc. This work describes the capsule design, neutronic and thermal analyses, and flow testing that were performed to support the qualification of this new irradiation vehicle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Assessment of Te as a U-Zr fuel additive to mitigate fuel-cladding chemical interactions

Tellurium was investigated as a potential additive to metallic fuel (e.g. U-10Zr) to mitigate fuel-cladding chemical interaction (FCCI). A primary cause of FCCI is the lanthanide fission products migrating to the fuel periphery and interacting with cladding, Te therefore was used to bind the lanthanides into stable compounds in the U-10Zr. The present work investigates the microstructures of as-cast and annealed U-10Zr-4Te and U-10Zr-4Te-4Ce alloys (weight percent) with scanning electron microscope and transmission electron microscopy. In conclusion, tellurium was found to bind Ce and form CeTe compound, which will mitigate lanthanides migration to the cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Hydrothermal corrosion of 2nd generation FeCrAl alloys for accident tolerant fuel cladding

As part of an effort to develop advanced fuel cladding for use in light water reactors, several FeCrAl alloys have been developed at Oak Ridge National Laboratory. A second generation of these alloys bearing additional alloying elements over the previously tested model alloys were tested in boiling water reactor (BWR) conditions to determine their resistance to hydrothermal corrosion. Coupons with alloy compositions Fe–10Cr–6Al–2Mo, Fe–13Cr–5Al–2Mo, Fe–13Cr–6Al–2Mo, Fe–13Cr–7Al–2Mo, and Fe–13Cr–5Al–2Mo–1Nb were tested in normal water chemistry (NWC) and hydrogen water chemistry (HWC) for 9 months in continuously refreshing autoclaves. Commercial FeCrAl alloy APMT and Zircaloy-2 were also tested for comparison. Among samples exposed to HWC, Fe–13Cr–7Al–2Mo performed worst with an average mass loss of 1.3 mg/cm2 over 9 months. This mass loss represents an estimated thickness loss of approximately 60 μm over 6 years. Samples exposed to NWC had very small mass losses of less than 0.15 mg/cm2 or mass gains up to 0.05 mg/cm2. Based on the results of this testing, the 2nd generation FeCrAl alloys tested exhibit low wall thickness loss and are suitable in terms of corrosion resistance for use as LWR cladding in BWR-HWC and BWR-NWC under normal operating conditions.

Raiman, Stephen↗
DOE OSTI2018

Demonstration of Chemical Decladding and Disposition Options for Non-Al Clad Fuels in H-Canyon

Aqueous disposition of non-Al clad fuels was studied in order to determine if H-Canyon would be a viable disposition option for this type of L-Basin fuel. A review of the L-Basin fuel types and its condition was done suggesting two major cladding types, Zircaloy and 304L stainless steel needed to be studied. Dissolution flowsheets options were reviewed. Sulfex and Zirflex processes were chosen as best for H-Canyon integration. Both process flowsheets were demonstrated on a lab scale to determine decladding rate, and off-gas and waste production. The resulting data suggested decladding in H-Canyon was an option but dissolver corrosion as well as off-gas and waste generation needed to be addressed. Small scale dissolution in a hot cell of smaller damaged fuel is a viable option with the flowsheets tested.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Characterization of the Chlorination Products of Zirconium Alloys used for Nuclear Fuel Cladding

The United States produces over 2000 metric tons of used nuclear fuel material per year, of which 25% is irradiated fuel cladding, and it is expected to increase for the next 50 years. Previous simulation studies have shown that in the irradiated cladding, the major activation products found are {sup 125}Sb, {sup 60}Co, and {sup 99m}Nb, and the major fission product found is {sup 137}Cs. These radioisotopes make it impossible to treat the cladding as low level waste. A chlorination process has been proposed for zirconium reprocessing as it has been proven more efficient than previous mechanical processes, redox processes, and other chemical methods (e.g. iodination). In the chlorination process, zirconium alloy cladding reacts with Cl{sub 2} at 350- 380 deg. C to produce mainly ZrCl{sub 4}. The zirconium tetrachloride has a relatively low sublimation point (331 deg. C) that allows its separation from other products, and so it can be treated as low level waste for disposal or recycling. However, previous work has observed element impurities (iron, tin, chromium, cesium, antimony, and niobium) in the recovered ZrCl{sub 4}, meaning that additional steps are necessary in a chlorination process previous to the reduction of ZrCl, to metallic Zr. The emphasis of this research summary is on the progress that has been made to isolate and characterize the impurities produced after chlorination of three previously characterized non-irradiated zirconium alloys (Zircaloy-2, Zircaloy-4, and a commercial Zr-Nb alloy) widely used as cladding materials.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Metallographic examinations and hydrogen measurements of high-burnup spent nuclear fuel cladding

In the US, commercial spent nuclear fuel (SNF) is transferred to interim dry storage casks where it will be stored for decades awaiting transport to a consolidated interim storage facility or a geologic repository. Because the fuel rod cladding is the first barrier against any radioactive material release, understanding the behavior of SNF cladding, particularly at high burnup (HBU), in dry storage conditions is crucial to safely store and transport the spent fuel. In this study, a series of metallographic examinations and cladding hydrogen measurements were conducted on HBU SNF cladding at Oak Ridge National Laboratory as a part of the High Burnup Spent Fuel Data Project, which is sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE). Here, t o investigate the effect of simulated drying conditions on the cladding, three as-received fuel rods with different cladding materials—M5, ZIRLO, and Zircaloy-4—were heated to 400°C and then slow-cooled to room temperature. The pellet and cladding were then qualitatively and quantitatively analyzed and compared in terms of pellet crack morphology, HBU rim, waterside oxide, cladding H, and cladding hydride morphologies. This paper presents and discusses the results of these analyses in detail.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2025

Implementing Fuel Cladding Bonding and Assessing its Impact on Axial Gas Communication

Fuel rods irradiated in light-water reactors to burnup values above 45 MWd/kgU are subject to the formation of a chemical and mechanical bond between the fuel and cladding upon gap closure. The formation of the bond subsequently inhibits the ability of fission gases released from the fuel to flow freely to the plenum of the rod. The flowing of gases within fuel rods is referred to as axial gas communication. During transients, such as loss of coolant accidents, the bond may influence cladding deformation prior to breaking. Upon bond breakage, gases are able to more freely communicate to the lower pressure regions of the rod. The impact of bonding on gas communication and the ballooning behavior of the cladding during a loss of coolant accident is of interest to the nuclear industry in support of burnup extension for the existing light-water reactor fleet. In this report, a model to capture the effects of fuel cladding bonding in the BISON fuel performance code is presented. The theory of the model along with implementation testing is provided. A summary of a previously developed axial gas communication is given to set the stage for how the two models may be coupled together. A full-length pressurized-water reactor fuel rod demonstration is highlighted to evaluate the impact of including bonding on axial gas communication calculations using a preliminary coupling methodology. An overview of the next steps regarding the modeling of bonding, axial gas communication, and a more tightly coupled framework is also provided.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Effect of surface oxides on tritium entrance and permeation in FeCrAl alloys for nuclear fuel cladding: a review

Abstract Iron-chromium-aluminum (FeCrAl) alloys are being considered for the cladding of uranium dioxide fuel in light water reactors (LWRs). FeCrAl alloys have good mechanical properties at temperatures of 300 °C and higher, and have superlative resistance to attack by steam at temperatures of up to 1000 °C and higher. A concern has been raised that the use of FeCrAl for cladding would result in a higher content of tritium in the reactor coolant as compared with the current system where the cladding is a zirconium based alloy. This review shows that the flux of tritium from the fuel rod cavities to the coolant across the fuel cladding wall will be greatly reduced by the presence of oxides on the surface of the cladding. The review of current literature and permeation data show that (a) protective oxides are expected to be present on both sides of the FeCrAl cladding, and (b) depending on the characteristics of these oxide layers it is reasonable to expect about two–three orders of magnitude reduction in tritium permeation, relative to the permeation response in clean, unoxidized condition for FeCrAl steels of interest, around 277 °C–377 °C temperatures.

Electrochemistry↗
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↗
DOE OSTI2018

Development of cone-wedge-ring-expansion test to evaluate the tensile HOOP properties of nuclear fuel cladding

Here, we report the development of a cone wedge ring expansion test method to determine the tensile hoop properties of irradiated nuclear fuel cladding in a hot cell. A closed-form solution was developed to explicitly relate the hoop stress σ θ to the axial piston loading using the scale factor χ and the shear stress correction factor α, where the χ and α factors are used to convert the applied piston load and the radial dilatation of the ring into the tensile hoop stress-strain curve. The applicability of this new test method was demonstrated by successfully evaluating ring specimens of Zr-4 alloy to failure. The ring specimens deformed uniformly in the axial and circumferential directions until onset of necking. The χ and α factors were successfully determined by using a finite-element stress analysis that benchmarked with applied test load and total work to account for frictional forces and the associated energies dissipation between the different sliding components of the testing system. The tensile hoop strength values obtained using the proposed cone-wedge-ring-expansion test method were 10% higher than values obtained from the tensile evaluation of the cladding material along the axial direction. In conclusion, the variability in results was found to be within the range reported in the literature.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A pulse-controlled modified-burst test instrument for accident-tolerant fuel cladding

Pellet-cladding mechanical interaction due to thermal expansion of nuclear fuel pellets during a reactivity-initiated accident (RIA) is a potential mechanism for failure of nuclear fuel cladding. To investigate the mechanical behavior of cladding during an RIA, we developed a mechanical pulse-controlled modified burst test instrument that simulates transient events with a pulse width from 10 to 300 ms. This paper includes validation tests of unirradiated and prehydrided ZIRLO cladding tubes. A ZIRLO cladding sample with a hydrogen content of 168 wt. ppm showed ductile behavior and failed at the maximum limits of the test setup with hoop strain to failure greater than 9.2%. ZIRLO samples showed high resistance to failure even at very high hydrogen contents (1,466 wt. ppm). When the hydrogen content was increased to 1,554 wt. ppm, brittle-like behavior was observed at a hoop strain of 2.5%. Preliminary scoping tests at room temperature with FeCrAl tubes were conducted to imitate the pulse behavior of transient test reactors during integral tests. The preliminary FeCrAl tests are informative from the perspective of characterizing the test rig and supporting the design of integral tests for current and potentially accident tolerant cladding materials.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Capabilities to improve corrosion resistance of fuel claddings by using powerful laser and plasma sources

The treatment conditions of fuel claddings of the E110 alloy by using powerful UV or IR laser radiation, which lead to the increase in the corrosion resistance at the high-temperature (T = 1100°C) oxidation simulating a loss-of-coolant accident, are determined. The possibility of the complete suppression of corrosion under these conditions by using pulsed laser deposition of a Cr layer is demonstrated. The behavior of protective coatings of Al, Al{sub 2}O{sub 3}, and Cr planted on steel EP823 by pulsed laser deposition, which is planned to be used in the BREST-OD-300, is studied. The methods of the almost complete suppression of corrosion in liquid lead to the temperature of 720°C are shown.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Improvement of mechanistic fuel-cladding chemical interaction modeling in BISON

This report describes work performed during FY2024 under the auspices of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to inform and improve mechanistic models of fuel-cladding chemical interaction (FCCI) in metallic fuel. For fuel-side FCCI, atomistic simulations were performed to determine the diffusivity of iron (Fe) in the $\alpha$ and $\gamma$ phases of uranium (U). A model of liquid penetration of cladding due to melting of the fuel-side FCCI region was updated to account for the finite size of the FCCI region, and the model was validated through comparison with tests performed in the Fuel Behavior Test Apparatus (FBTA). For cladding wastage formation, a reduced-order model was improved by comparison with a multi-scale mechanistic model to better quantify the ROM parameters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Characterization of the Hydrothermal Corrosion Behavior of Ceramics for Accident Tolerant Fuel Cladding

Accident-tolerant fuel (ATF) is an increasingly important research topic for the nuclear industry, and ceramics such as SiC are strong contenders for deployment as ATF cladding. The hydrothermal corrosion characteristics of SiC and Al2O3 were investigated via constantly-refreshing autoclave corrosion and post exposure characterization. Four different types of chemical vapor deposited (CVD) SiC specimens were examined (two with high electrical resistance, one with low electrical resistance, and a single crystal 4H structural variant). Al2O3 specimens were prepared in single crystal and polycrystalline states. PWR primary water, BWR–HWC, and BWR–NWC environments were maintained throughout the experiments. Characterization conducted using SEM and EDS was used to determine factors affecting corrosion rates and susceptibility to grain boundary attack in each water chemistry condition. Raman spectroscopy was also used to determine chemical variation of the surface with corrosion. Grain boundary attack was found to be significant for both alumina and SiC polycrystalline variants.

Doyle, Peter↗
DOE OSTI2022

Microstructural and nanomechanical studies of PVD Cr coatings on SiC for LWR fuel cladding applications

The microstructure and nanomechanical properties of pure Cr coatings deposited on SiC with various physical vapor deposition (PVD) methods have been investigated for the mitigation of hydrothermal corrosion of SiC-SiC f (SiC fiber-reinforced SiC matrix) composite fuel cladding in light water reactors (LWR). Cr coatings (4–7 μm thick) were deposited on unbiased SiC substrates using six variants of magnetron sputtering processes: (i) standard (S-DCMS), (ii) pulsed (P-DCMS), (iii) ion-assisted (I-DCMS), and (iv) pulsed ion-assisted direct current magnetron sputtering (PI-DCMS), (v) high-power impulse magnetron sputtering (HiPIMS), and (vi) bipolar HiPIMS (B-HiPIMS). Microstructural characterization and nanoindentation testing were used to evaluate effects of deposition technique on coating microstructure and nanomechanical properties. The application of positive reverse pulses to the sputter target induced coarsening of the columnar grain structure, while ion bombardment of substrates during deposition promoted densification of the coating microstructure. Both effects are exemplified in B-HiPIMS deposition, resulting in a high-density microstructure with compressive residual stress. Material pile-up around nano-indents and plastic work during nanoindentation were analyzed to understand deformation behavior of the coatings. Finally, the results suggest the B-HiPIMS process to be among the promising methods for the deposition of Cr coatings on SiC-SiC f for LWR cladding application.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Electron microscopy characterization of the fuel-cladding interaction in medium burnup annular fast reactor MOX

In this work, we present electron microscopy data focused on the fuel-cladding interaction layer in annular fast reactor MOX with HT-9 cladding at medium burnup. In agreement with previous literature data, the volatile fission products Cs, Te and I have migrated radially into the extreme fuel periphery and partially interacted with the cladding. The accumulation of Cs has occurred in the outermost rim of the fuel pellet, where grain recrystallization has also been observed. Significant amounts of Pd have been found in the interaction zone, particularly in the sample taken from the upper half of the fissile column where the cladding temperatures are higher. At this axial location, Cr has been enriched at the cladding inner surface and diffused into the fuel. Furthermore, chromium remains mainly in metallic form, but locally formed oxides. The fission products Cs, Te and I are found, with variable composition, in form of nanocrystalline regions dispersed in the metallic Cr-rich layer. The morphology and chemical characteristics of the layer suggest a non-oxidative corrosion mechanism as principal cladding degradation phenomenon occurring in this sample, with local onset of Cr oxidation within the nanocrystalline precipitates.

36 MATERIALS SCIENCE↗
DOE OSTI2023

In situ high-temperature 3D imaging of the damage evolution in a SiC nuclear fuel cladding material

Silicon carbide (SiC)-based nuclear fission fuel rod cladding has been considered as one of the possible designs for accident tolerant fuels. It is in the form of a SiC fibre reinforced SiC matrix composite tube (SiC f -SiC m ) with monolithic SiC outer and/or inner coating layers. This study focuses on the deformation and fracture processes in this material using in situ X-ray micro-computed tomography (XCT) at room temperature (RT) and 1200 °C in an inert gas environment in a C-ring compression loading configuration. Prior to testing, local properties and residual stresses were characterised using nanoindentation and Raman spectroscopy since they can impact the mechanical behaviour of the material. The 3D strain distribution, crack formation and propagation processes including the toughening mechanisms (e.g., crack deflection, micro-cracking, crack bridging and bifurcation) are investigated in the coating and underlying composites at RT and 1200 °C. There is no particular sequence which toughening mechanism occurs first – this is very different from the conventional toughening theory in ceramic-matrix composites under uniaxial tension loading. Indeed, no evidence of fibre pull-out or fibre fracture was observed in this SiC f -SiC m nuclear cladding material in the current C-ring compression configuration. The correlation between local measurements and bulk mechanical behaviour are discussed.

3D imaging↗
DOE OSTI2019

Radiation response of Ti 2 AlC MAX phase coated Zircaloy-4 for accident tolerant fuel cladding

A Ti 2 AlC MAX phase material was deposited on Zircaloy-4 substrate using the cold spray process, with the goal of enhancing the accident tolerance of zirconium-alloy fuel cladding in light water reactors. The samples were annealed up to 923 K to study the intermetallic phase formation between the coating and Zircaloy-4 substrate. Three interface compounds, ZrAl 2 , ZrAl, and Zr 3 Al, were identified and the activation energy of the interfacial intermetallic compound growth was determined. The radiation response of the coating material and interfacial intermetallic compounds after 3.5 MeV Zr 2+ ion irradiation was investigated by transmission electron microscopy and nanoindentation. Ion irradiation disrupted the nanolamellar structure of the Ti 2 AlC coating while amorphization was observed some of the intermetallic compounds. Furthermore, nanoindentation revealed overall hardening of the Ti 2 AlC coating and Zr 3 Al phase, with no significant change observed in the ZrAl 2 and ZrAl phases.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗