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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

NON-DESTRUCTIVE POST IRRADIATION EXAMINATION OF FIRST CYCLE ACCIDENT TOLERANT AND ADVANCED ZIRCONIUM ALLOY HIGH BURNUP FUEL RODS

Post irradiation examination (PIE) of accident tolerant fuel concepts and high burnup uranium dioxide advanced zirconium alloy clad fuel contribute to near term nuclear industry goals of enhancing light water reactor safety and economics. Westinghouse is developing chromium coated zirconium alloy cladding as an accident tolerant fuel option for current pressurized water reactors. Examination of high burnup fuel further augments the technical foundation for extending the peak rod average burnup limits beyond the current regulatory limit of 62 MWd/kgU. Seven fuel rods were received at the Oak Ridge National Laboratory Irradiated Fuel Examination Facility. This includes three chromium coated zirconium alloy fuel rods and four high burnup advanced zirconium alloy fuel rods. Six of the seven rods were fueled with standard uranium dioxide while one of the chromium coated rods contained Westinghouse’s ADOPT fuel. The overall hot-cell PIE plan and some PIE results have been presented previously. This paper focuses on the additional non-destructive examination and fission gas release. Highlights from visual examination of these rods are presented. The chromium coating showed little to no visual change following irradiation. Observations of the axial burnup trends in the fuel rods are derived from full rod axial gamma-ray scans with additional observations related to local fission product migration. The axial dimensional changes of the seven rods are evaluated. As expected, creep down had not yet occurred in the chromium coated rods, and the maximum diameter strain in the high burnup advanced zirconium alloy cladding was less than 0.5%. Finally, the steady state fission gas release for these rods was measured ranging from 4.1% to 17.6% which matched expectation from literature. These fuel rods now await further destructive examination.

Harp, Jason↗
DOE OSTI2013

Synthesis of zirconium oxide nanoparticle by sol-gel technique

Zirconium oxide nanoparticle is synthesized using sol-gel technique. Various mole ratio of ammonia solution and nitric acid relative to zirconium propoxide is added in the reaction to study the effect on the crystallinity and particle size on zirconium oxide particle. Zirconium oxide synthesized with nitric acid have the smallest particle size under FESEM image and show the increasing formation of crystalline tetragonal phase under XRD diffractogram.

36 MATERIALS SCIENCE↗
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 OSTI2017

Progressing Zirconium-Alloy Corrosion Models Using Synchrotron XANES

The corrosion and hydrogen pickup of in-reactor zirconium-based cladding is currently limiting the maximum fuel burnup in light-water reactors. Since the oxidation rate and hydrogen pickup fraction of zirconium alloys vary significantly as a function of exposure time, burnup, and alloy composition, it is critical to better understand the underlying mechanisms to model and predict corrosion behavior. Following the analysis of ~500 autoclave coupons, a physically based zirconium-alloy corrosion model founded on first principles, named “Coupled Current Charge Compensation (C4)”, has been developed. The model reproduces the differences in oxidation kinetics and hydrogen pickup between different zirconium alloys, such as Zr-Nb and Zircaloy-4. Since oxidized solute elements affect the corrosion process through a space-charge compensation mechanism, synchrotron nano-beam X-ray Absorption Near-Edge Spectroscopy has been performed on multiple oxidized Zr-Nb alloys to determine the oxidation-state profile of niobium in the oxide layer. The results inform the C4 model and the associated hydrogen pickup fraction.

C4 Model↗
DOE OSTI2018

Mechanism and Conditions for Continuous Hydride Layer Formation in Zirconium Cladding

Local hydrogenation of zirconium fuel-element cladding, resulting in the formation of continuous hydrides, as a rule occurs when coolant or other hydrogen-containing compound finds its way beneath the fuel-element cladding irradiated in the core of a water moderated and cooled reactor. In this article, the set of processes occurring during hydrogenation of zirconium cladding is analyzed, and estimates are given for the parameters of these processes. The conditions and mechanism of formation of a continuous hydride over the thickness of zirconium cladding are established. It is shown that the duration of the nucleation of the hydride is determined primarily by the time to local destruction of the oxide film on the inner surface of the zirconium cladding of the fuel elements. The shortest possible time for continuous hydride to grow through the thickness of fuel-element cladding is determined.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Separation of zirconium and hafnium from early actinides and rare earth elements with eichrom’s pb resin in HCl

The separation of zirconium and hafnium isotopes from the early actinides and rare earth elements (REE) with Eichrom’s Pb resin has been studied. Batch studies were performed to characterize the behavior of actinium, thorium, zirconium, hafnium, lutetium, and yttrium on Pb resin from HCl solutions (0.001 M to 11 M). The early actinides and REE had no affinity for the resin at any concentration of HCl, but zirconium and hafnium showed a moderate uptake at high concentrations of HCl with a maximum extraction at 11 M HCl. Several column separations were tested, including with only tracer isotopes and with mass. Rapid, simple separations of zirconium from actinium, thorium, protactinium, and the REE with high yields and low elution volumes are presented with applications for tracer isotope production and fission product separations. Furthermore, the resin is less suitable for hafnium separations as hafnium tends to bleed off the resin even at high concentrations of HCl.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2018

Aluminum Ion Beam Treatment of Zirconium Ceramics

The paper presents the radiation and thermal treatment of zirconium ceramics with high-energy Al ion beams generated at an accelerating voltage of 1.5 kV, which modifies the structure and electrophysical properties of zirconium ceramics. Compact powder and ceramic samples are used for the radiation and thermal treatment performed at 1123–1173 K. The surface treatment of compact powders leads to the increase in the grain size, whereas the surface of ceramic samples turns black and electrically conductive in depth. This is because the change in the oxygen stoichiometry of zirconium ceramics. Air annealing of treated ceramics returns the sample to the initial state. The phase composition, microhardness and density of ceramic samples display no changes after the radiation and thermal treatment. Under the experimental conditions, the diffusion of aluminum ions in the surface layer is not observed. It is found that the ion beam treatment leads to the decrease in aluminum-containing impurity in the surface layers of zirconium ceramics.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2019

Role of Titanium, Carbon, Boron, and Zirconium in Carbide and Porosity Formation during Equiaxed Solidification of Nickel-Based Superalloys

This work analyses the effect of Ti in the MC carbide formation and also in the γ/γ´ eutectic promotion in two nickel-based alloys melted under vacuum conditions and solidified into ceramic molds. The role of boron, carbon and zirconium on the porosity level after solidification of both alloys was also investigated; since the alloys were vacuum melted, it is assumed that the observed porosity is due to micro-shrinkage. The chemical compositions for two alloys were: 20.2Cr-13.7Co-4.78Mo-1.19Al-0.69Ti-0.08C-2.34Fe for Alloy 1, and 20.5Cr-15.6Co5.03Mo-1.57Al-3.42Ti-0.15C-1.64Fe-0.076Zr-0.01%B for Alloy 2. Results showed a higher volume fraction of MC carbides and the γ/γ´ eutectic for the Alloy 2 with higher titanium content since this element is a carbide and γ/γ´ eutectic promoter. The porosity level in the Alloy 2, with higher amounts of carbon, boron and zirconium was considerably lower than that for the Alloy 1. These results are discussed in terms of different phenomena caused by the segregating effects of boron, carbon and zirconium towards the interdendritic regions and the formation of zirconium carbides that produce an expansion in volume when they are formed at the end of the solidification process.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Intercalation chemistry of zirconium 4-sulfophenylphosphonate

Zirconium 4-sulfophenylphosphonate is a layered material which can be employed as a host for the intercalation reactions with basic molecules. A wide range of organic compounds were chosen to represent intercalation ability of zirconium 4-sulfophenylphosphonate. These were a series of alkylamines from methylamine to dodecylamine, 1,4-phenylenediamine, p-toluidine, 1,8-diaminonaphthalene, 1-aminopyrene, imidazole, pyridine, 4,4′-bipyridine, poly(ethylene imine), and a series of amino acids from glycine to 6-aminocaproic acid. The prepared compounds were characterized by powder X-ray diffraction, thermogravimetry analysis and IR spectroscopy and probable arrangement of the guest molecules in the interlayer space of the host is proposed based on the interlayer distance of the prepared intercalates and amount of the intercalated guest molecules. - Graphical abstract: Nitrogen-containing organic compounds can be intercalated into the interlayer space of zirconium 4-sulfophenylphosphonate. - Highlights: • Zirconium 4-sulfophenylphosphonate was examined as a host material in intercalation chemistry. • A wide range of nitrogen-containing organic compounds were intercalated. • Possible arrangement of the intercalated species is described.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Trap state passivation improved hot-carrier instability by zirconium-doping in hafnium oxide in a nanoscale n-metal-oxide semiconductor-field effect transistors with high-k/metal gate

This work investigates the effect on hot carrier degradation (HCD) of doping zirconium into the hafnium oxide high-k layer in the nanoscale high-k/metal gate n-channel metal-oxide-semiconductor field-effect-transistors. Previous n-metal-oxide semiconductor-field effect transistor studies demonstrated that zirconium-doped hafnium oxide reduces charge trapping and improves positive bias temperature instability. In this work, a clear reduction in HCD is observed with zirconium-doped hafnium oxide because channel hot electron (CHE) trapping in pre-existing high-k bulk defects is the main degradation mechanism. However, this reduced HCD became ineffective at ultra-low temperature, since CHE traps in the deeper bulk defects at ultra-low temperature, while zirconium-doping only passivates shallow bulk defects.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2018

Mutual influence of hydrogen and vacancies in α-zirconium on the energy of their interaction with metal

We presents a first-principles investigation on the interaction energy of hydrogen and vacancies with α-zirconium. It is established that the presence of vacancies in zirconium increases the hydrogen–metal binding energy; the presence of hydrogen in the zirconium lattice reduces the vacancy formation energy. It is shown that hydrogen and vacancies in zirconium form complexes that considerably distort the metal lattice. The increase in the covalency degree of the metal–metal and hydrogen–metal bounds is observed in the vicinity of these complexes.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗
DOE OSTI2014

Investigation of Electrochemical Recovery of Zirconium from Spent Nuclear Fuels

This project uses both modeling and experimental studies to design optimal electrochemical technology methods for recovery of zirconium from used nuclear fuel rods for more effective waste management. The objectives are to provide a means of efficiently separating zirconium into metallic high-level waste forms and to support development of a process for decontamination of zircaloy hulls to enable their disposal as low- and intermediate-level waste. Modeling work includes extension of a 3D model previously developed by Seoul National University for uranium electrorefining by adding the ability to predict zirconium behavior. Experimental validation activities include tests for recovery of zirconium from molten salt solutions and aqueous tests using surrogate materials. *This is a summary of the FY 2013 progress for I-NERI project # 2010-001-K provided to the I-NERI office.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Isothermal oxidation of zirconium using steam

Oxidation of zirconium and its alloys under any atmosphere is generally characterised by a protective and non-protective stage. During the protective stage, the kinetics of oxidation are quasi-parabolic, while the oxide scale formed on the surface of the metal is black, compact, and protective in nature. For the non-protective stage, the oxidation rate becomes linear, while the protective oxide film turns into a deteriorated white scale. Breakaway is known as the transition point from protective to non-protective stage. The actual mechanism leading to breakaway of zirconium is still a matter of controversy, however some of the most common factors affecting breakaway are the following: - Impurities on the metal (Carbon, Oxygen, Nitrogen); - Oxidizing atmosphere (Presence of H{sub 2}O (g) or N{sub 2} (g)); - Sample preparation (Chemically etched or polished); - Temperature (LWR, LOCA). This research investigates the effect of temperature on 'breakaway' when zirconium is oxidized isothermally in steam for 60 min at 800 and 900 deg. C using thermogravimetric and optical microscopic imaging techniques. 1. Steam isothermal oxidations were performed on zirconium at 800 and 900 deg. C using a Thermogravimetric Analyzer (TGA). 2. Log-log plots of mass gain vs time were obtained from thermogravimetric results to calculate the reaction orders (n) and reaction rate constants (K) during oxidation. 3. Results of the kinetics of reaction were correlated with optical microscopic images form Light Microscope (LM) and Scanning Electron Microscope (SEM). Breakaway depends very strongly on temperature. At 800 deg. C, the weight gain followed a cubic-parabolic behaviour with evidence of a protective oxide film. On the other hand, oxidation at 900 deg. C showed signs of breakaway since the oxide film was cracked and oxidation rate was almost linear throughout the reaction.

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

Alloying effect of Nb and Sn on the Zirconium alloy fuel claddings behavior at high temperature oxidation in steam - 6.1 STP-2016-0072

As the material of fuel claddings for water cooled reactors most commonly used binary and multicomponent zirconium alloys Zr-Nb (E110,M5), Zr-Sn-Fe (Zircaloy-2,-4) and Zr-Nb-Sn-Fe (E635, Zirlo) are used, which improvement extends, in particular, by varying their composition by Nb, Sn and Fe. At the same time, the importance is given to ensuring the safe operation of the fuel rods not only in normal conditions but also in emergency situations such as LOCA. In this paper the researches results of the studies of influence of alloying elements Nb, Sn (and Fe) on corrosion resistance and embrittlement at high temperature steam oxidation of fuel claddings based on alloys types Zr-xNb (x=1.0/2.5) and Zr-xNb-ySn-zFe (x=0.6/2.4; y=0.24/1.1; z=0.18/0.34) made with using zirconium sponge are presented. High temperature steam oxidation tests were carried out at temperatures of 1000, 1100 and 1200 deg. C with continuous measurement of the weight gain during the experiment and subsequent cooling in steam with rate ∼ 20 deg. C/s. The studies of kinetics of high-temperature oxidation in steam, structural-phase state changes, alloying elements distribution, absorbed hydrogen content and residual ductility after rapid cooling of the oxidized specimens were made. The difference in the oxidation kinetics of the materials studied was revealed, it decreases with the oxidation temperature increase. At the same time, an increase of the Sn content in the alloy influences on specimens 'breakaway' oxidation intensification at 1000 deg. C. It is shown that the mechanical properties of fuel claddings oxidized under the same conditions depend on their alloying composition, the hydrogen fraction absorbed by the specimen and formed by the oxidation ZrO2, a-Zr(O) and 'ex-β' layers structure, which in turn varies depending on the alloy composition. The increase of Nb, Sn and Fe content in the zirconium alloy leads to decrease in residual ductility of fuel claddings after high temperature steam oxidation. According to obtained results the most resistant to high-temperature steam oxidation and embrittlement at temperatures of 1000-1200 deg. C among the materials studied is Zr-1Nb alloy. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Computational Studies of Gas Phase Reactions of Zirconium Tetrachloride (ZrCl{sub 4}) and Metal Chloride Impurities

Zirconium-based cladding materials that surround used nuclear fuel (UNF) rods represent a significant fraction of the total amount of high level waste that must be disposed of in long term geological repositories. The decontamination, recovery and reuse of these materials would significantly reduce the high level waste associated with light water reactors and could provide a new source of hafnium free zirconium for the fabrication of new nuclear components. One part in the strategy to recycle zirconium from used cladding materials involves developing a computational modeling (DFT) protocol to understand the reactions between ZrCl{sub 4} and metal chloride impurities that may occur in the gas phase during sublimation. Several of the metal chloride species present are known to form Lewis acid/base adducts in the gas phase and development of purification protocols require investigation into the properties of these ZrCl{sub 4} contaminates.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Generation of gas-phase zirconium fluoroanions by electrospray of an ionic liquid

RATIONALE: When measuring extremely wide isotope ratios (= 1 x 109) accelerator mass spectrometry (AMS) is the instrument of choice, however it requires an anion for injection into the tandem accelerator. Since many elements do not have positive electronegativities they do not form stable negative atomic ions, and hence are not compatible for isotope ratio measurement using AMS. Thus new approaches for forming anions are sought; fluoroanions are particularly attractive because fluorine is monoisotopic, and thus will not have overlapping isobars with the isotope of interest. METHODS: An approach is described for making zirconium fluoroanions using the fluorinating ionic liquid (IL) 1-ethyl-3-methylimidazolium fluorohydrogenate, which was used to generate abundant [ZrF5-] using electrospray ionization. The IL was dissolved in acetonitrile, combined with a dilute solution of either Zr4+ or ZrO2+, and then electrosprayed. Mass analysis and collision induced dissociation were conducted using a time-of-flight mass spectrometer. Cluster structures were predicted using density functional theory calculations. RESULTS: The fluorohydrogenate IL solutions generated abundant [ZrF5-] starting from solutions of both Zr4+ and ZrO2+. The mass spectra also contained IL-bearing cluster ions, whose compositions indicated the presence of [ZrF6]2- in solution, a conclusion supported by the structural calculations. Rinsing out the zirconium-IL solution with acetonitrile decreased the IL clusters, but enhanced [ZrF5]-, which was sorbed by the polymeric electrospray supply capillary, and then released upon rinsing. This reduced the ion background in the mass spectrum. CONCLUSIONS: The fluorohydrogenate-IL solutions are a facile way to form zirconium fluoroanions in the gas phase using electrospray. The approach has potential as a source of fluoroanions for injection into an AMS, which would enable high-sensitivity measurement of minor zirconium isotopes, and benefits from the absence of overlapping isobars caused by the charge carrier (i.e., the monoisotopic fluorine atoms).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2025

Experimental confirmation of first-principles thermal conductivity in Zirconium-doped ThO 2

The degradation of thermal conductivity in advanced nuclear fuels due to the accumulation of fission products and irradiation-induced defects is inevitable, and must be considered as part of safety and efficiency analyses of nuclear reactors. Here, this study examines the thermal conductivity of a zirconium-doped ThO 2 crystal, synthesized via the hydrothermal method using a spatial domain thermoreflectance technique. Zirconium is one of the soluble fission products in oxide fuels that can effectively scatter heat-carrying phonons in the crystalline lattice of fuel. Thus, thermal property measurements of zirconium-doped ThO 2 single crystals provide insights into the effects of substitutional zirconium doping, isolated from extrinsic factors such as grain boundary scattering. The experimental results are compared with first-principles calculations of the lattice thermal conductivity of ThO 2 , employing an iterative solution of the Peierls-Boltzmann transport equation. Additionally, the non-perturbative Green's function methodology is utilized to compute phonon-point defect scattering rates, accounting for local distortions around point defects, including mass difference changes, interatomic force constants, and structural relaxation. The congruence between the predicted results from first-principles calculations and the measured temperature-dependent thermal conductivity validates the computational methodology. Furthermore, the methodologies employed in this study enable systematic investigations of thermal conductivity reduction by fission products, potentially leading to the development of more accurate fuel performance codes.

36 - MATERIALS SCIENCE↗
DOE OSTI2025

In situ observation of irradiation-induced enhancement to the desorption pressure of zirconium hydride in a nuclear reactor

We quantify the effect of a nuclear-reactor environment on the hydrogen isotope equilibrium vapor pressure over pure zirconium and zirconium hydride. A vacuum-sealed capsule containing a zirconium foil with 6 atom% deuterium was irradiated at a neutron flux of ~10 14 cm -2 s -1 at the University of Missouri Research Reactor (MURR). The internal stainless-steel (SS) sample holder acted as the heat source via gamma absorption. To measure low desorption pressures in a high-flux environment, we developed a method to transduce pressure from the measured sample temperature during irradiation, calibrating with known deuterium pressures in unirradiated capsules at various heating powers using an internal filament-heated system designed to mimic irradiation-induced heating. Our temperature-pressure transduction method operates similarly to a Pirani or thermocouple pressure gauge. The in-reactor measurements revealed a roughly 4-fold enhancement in desorption pressure after only 6 h of irradiation (~2 × 10 18 cm -2 neutron fluence) compared to thermal desorption in control experiments, indicating a nonthermal contribution from neutron irradiation. The slower temperature/pressure stabilization rate in the reactor suggests that desorption pressure enhancement increases with neutron fluence. Further, this enhancement signifies increased solubility of hydrogen isotopes in zirconium during irradiation. We propose that high-energy neutron collisions with hydrogen isotopes in hydrides lead to their decomposition at lower temperatures, supersaturating the surrounding αZr lattice and resulting in higher desorption pressure, which continues to rise as more hydrides dissolve with increasing neutron fluence.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗