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At least 19 records
All sourcesNASA NTRSDOE OSTI
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 OSTI2022

Report summarizing demonstration of thermal conductivity measurement methods for coated zirconium cladding

Thermal diffusivity measurements on Zirconium-based cladding materials have historically been a challenge due to the difficulty to measure on specimens with curved geometries, including nuclear grade Zircaloy cladding materials. In this work, we first used Laser Flash Analysis methods to measure the thermal diffusivity of Zircaloy tubes, which show good agreement with Zircaloy plates in this work, as well as previously published data. The consistent results proved the applicability of the Laser Flash Analysis setup for investigating thermal diffusivity of Zircaloy tubes. We further investigate the hydrogen effect on thermal diffusivity of Zircaloy tubes. Hydrogen plays significant roles in temperature dependent thermal diffusivity trends of Zircaloy tube. For low hydrogen concentration, where hydride dissolved into single phase α-Zr at higher temperature, thermal diffusivity is relatively unchanged at low temperature at two phase regions (α-Zr and δ-hydride), and it increases as a function of temperature at single phase α-Zr region. For high hydrogen concentration, where phase transformation (α-Zr + δ-hydride → α-Zr + β-Zr) occurs at 567 °C, thermal diffusivity decreases as a function of temperature below 567 °C before the transformation (α-Zr and δ-hydride), while increase at higher temperature after the transformation (α-Zr and β-Zr).

36 MATERIALS SCIENCE↗
DOE OSTI2016

Assessment of wear coefficients of nuclear zirconium claddings without and with pre-oxidation

In the cores of pressurized water nuclear reactors, water-flow induced vibration is known to cause claddings on the fuel rods to rub against their supporting grids. Such grid-to-rod-fretting (GTRF) may lead to fretting wear-through and the leakage of radioactive species. The surfaces of actual zirconium alloy claddings in a reactor are inevitably oxidized in the high-temperature pressurized water, and some claddings are even pre-oxidized. As a result, the wear process of the surface oxide film is expected to be quite different from the zirconium alloy substrate. In this paper, we attempt to measure the wear coefficients of zirconium claddings without and with pre-oxidation rubbing against grid samples using a bench-scale fretting tribometer. Results suggest that the volumetric wear coefficient of the pre-oxidized cladding is 50 to 200 times lower than that of the untreated cladding. In terms of the linear rate of wear depth, the pre-oxidized alloy wears about 15 times more slowly than the untreated cladding. Finally, fitted with the experimentally-determined wear rates, a stage-wise GTRF engineering wear model demonstrates good agreement with in-reactor experience in predicting the trend of cladding lives.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Impact of Coating Defects on Performance of Coated Zirconium Cladding

Research on accident tolerant fuels (ATF) has started after the Fukushima accident [1–3]. While efforts have been expended on both fuel and cladding ATF concepts, the bulk of work has been devoted to improved cladding. The overarching goal of these approaches is to extend the coping time available during a severe accident before any event would result in release of radioactivity to the public. The most basic ATF cladding concept is obtained by applying a thin coating of highly corrosion-resistant material on the surface of a licensed zirconium cladding alloy. This thin coating is intended to not interfere with the neutronic or mechanical performance of the base cladding under normal operating conditions. Different coating materials, thicknesses, coating processes, process parameters, and testing methods have an impact on the microstructure and mechanical properties and therefore on the results of the applied investigation methods. These challenges have motivated an initial focus on demonstrating that the presence of coatings do not perturb the critical performance benchmarks of uncoated material. Ongoing lead test assembly irradiations of coated zirconium concepts in commercial reactions is intended to establish baseline performance in this regard in the coming years.

36 MATERIALS SCIENCE↗
DOE OSTI2021

Deposition of integrated protective material into zirconium cladding for nuclear reactors by high-velocity thermal application

A zirconium alloy nuclear reactor cylindrical cladding has an inner Zr substrate surface, an outer volume of protective material, and an integrated middle volume of zirconium oxide, zirconium and protective material, where the protective material is applied by impaction at a velocity greater than 340 meters/second to provide the integrated middle volume resulting in structural integrity for the cladding.

Mazzoccoli, Jason P.↗
DOE OSTI2019

Purification of Zirconium Cladding Using a Chloride Volatility Process

The goal of this work is to investigate the thermochemistry of used fuel Zr cladding component chlorides as well as their activation products and residual radionuclide metal chlorides, in order to provide the DOE-NE community with fundamental data that will support the development of a chlorination process for the separation and purification of ZrCl 4 . Our studies will focus on the following elements: Zr, Sn, Fe, Cr, Sb, Nb and Cs; in the following sections, the term “chloride species” is applied to the following elements: Sn, Fe, Cr, Sb, Nb and Cs. Experimental and theoretical research will focus on the characterization, reactivity and volatility of the chloride species present in the ZrCl 4 obtained after chlorination of Zr cladding. The methods previously used for the purification of ZrCl 4 will be reexamined and the thermochemistry of the chloride species evaluated for each of the methods. Claddings of interest include Zircaloy-4, Zircaloy-2 and Zr-Nb alloys (Zirlo TM and Zr-2.5Nb). In this context, we propose five Tasks: Task 1: Computational study of the chemical form and thermochemical behavior of the chloride species produced after chlorination of Zr, Task 2: Characterization the chloride species obtained after chlorination of Zr cladding and Zr cladding components, Task 3 and Task 4: Study the thermochemical behavior of Zr cladding and Zr cladding complements in the presence of Cl 2 and HCl gas, and Task 5: Laboratory scale demonstrations of the removal of the impurity chloride species from ZrCl 4 .

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Development of Self-Healing Zirconium-Silicide Coatings for Improved Performance Zirconium-Alloy Fuel Cladding

Zirconium-alloy fuel claddings have been used successfully in Light Water Reactors (LWR) for over four decades. However, under high temperature accident conditions, zirconium-alloys fuel claddings exhibit profuse exothermic oxidation accompanied by release of hydrogen gas due to the reaction with water/steam. Additionally, the ZrO 2 layer can undergo monoclinic to tetragonal to cubic phase transformations at high temperatures which can induce stresses and cracking. These events were unfortunately borne out in the Fukushima-Daiichi accident in in Japan in 2011. In reaction to such accident, protective oxidation-resistant coatings for zirconium-alloy fuel claddings has been extensively investigated to enhance safety margins in accidents as well as fuel performance under normal operation conditions. Such surface modification could also beneficially affect fuel rod heat transfer characteristics. Zirconium-silicide, a candidate coating material, is particularly attractive because zirconium-silicide coating is expected to bond strongly to zirconium-alloy substrate. Intermetallic compound phases of zirconium-silicide have high melting points and oxidation of zirconium silicide produces highly corrosion resistant glassy zircon (ZrSiO 4 ) and silica (SiO 2 ) which possessing self-healing qualities. Given the long-term goal of developing such coatings for use with nuclear reactor fuel cladding, this work describes results of oxidation and corrosion behavior of bulk zirconium-silicide and fabrication of zirconium-silicide coatings on zirconium-alloy test flats, tube configurations, and SiC test flats. In addition, boiling heat transfer of these modified surfaces (including ZrSi 2 coating) during clad quenching experiments is discussed in detail.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

A novel protocol to recycle zirconium from zirconium alloy cladding from used nuclear fuel rods

A potential new zirconium recycling protocol has been demonstrated using unirradiated Zircaloy-based claddings and depleted uranium oxide for the chemical removal of zirconium alloy cladding in used nuclear fuel (UNF) rods from light water reactors. This protocol is based on the application of three new scientific findings. First, a new lower temperature chlorination reaction of zirconium with sulfur monochloride is described. Second, the high solubility of zirconium chloride in thionyl chloride is used to separate it from uranium oxide fuel and fission and activation contaminants. Finally, in the third step, the zirconium chloride is purified by simple recrystallization from thionyl chloride. Utilizing sulfur chloride solvents for a lower temperature liquid-based chemical digestion and purification of zirconium alloy claddings reduces technical complications experienced by current high temperature gas phase chlorination strategies, such as contamination of product streams. This novel protocol has been demonstrated on a 3–50 g scale of unirradiated zirconium alloy, with no significant change in the time required for complete chlorination (3–4 h). This zirconium chlorination protocol has also been performed in the presence of depleted uranium oxide pellets. The depleted fuel pellets do not affect zirconium chlorination, and the chlorination protocol does not chemically or physically alter the fuel pellets. Finally, a preliminary description of an industrial protocol to recycle nuclear grade zirconium from UNF rods is presented.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

White Paper Summary of 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding

This white paper recommends that ASTM International develop standards to address the potential impact of hydrides on the long term performance of irradiated zirconium alloys. The need for such standards was apparent during the 2nd ASTM International Workshop on Hydrides in Zirconium Alloy Cladding and Assembly Components, sponsored by ASTM International Committee C26.13 and held on June 10-12, 2014, in Jackson, Wyoming. The potentially adverse impacts of hydrogen and hydrides on the long term performance of irradiated zirconium-alloy cladding on used fuel were shown to depend on multiple factors such as alloy chemistry and processing, irradiation and post irradiation history, residual and applied stresses and stress states, and the service environment. These factors determine the hydrogen content and hydride morphology in the alloy, which, in turn, influence the response of the alloy to the thermo-mechanical conditions imposed (and anticipated) during storage, transport and disposal of used nuclear fuel. Workshop presentations and discussions showed that although hydrogen/hydride induced degradation of zirconium alloys may be of concern, the potential for occurrence and the extent of anticipated degradation vary throughout the nuclear industry because of the variations in hydrogen content, hydride morphology, alloy chemistry and irradiation conditions. The tools and techniques used to characterize hydrides and hydride morphologies and their impacts on material performance also vary. Such variations make site-to-site comparisons of test results and observations difficult. There is no consensus that a single material or system characteristic (e.g., reactor type, burnup, hydrogen content, end-of life stress, alloy type, drying temperature, etc.) is an effective predictor of material response during long term storage or of performance after long term storage. Multi-variable correlations made for one alloy may not represent the behavior of another alloy exposed to identical conditions and the material responses to thermo-mechanical exposures will be different depending on the materials and systems used. The discussions at the workshop showed several gaps in the standardization of processes and techniques necessary to assess the long term performance of irradiated zirconium alloy cladding during dry storage and transport. The development of, and adherence to, standards to help bridge these gaps will strengthen the technical basis for long term storage and post-storage operations, provide consistency across the nuclear industry, maximize the value of most observations, and enhance the understanding of behavioral differences among alloys. The need for, and potential benefits of, developing the recommended standards are illustrated in the various sections of this report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2021

Produce ZrH Moderator Material

This reports complete Milestone M2.2.3 Produce zirconium hydride moderator material. In this milestone, we report the successful fabrication of FeCrAl alloys cladded zirconium hydride moderator. Different types of crucible designs were presented, including the basic design with bar cladding material, basic design with tube cladding material, and crucible design with position holding function. Mo and FeCrAl were used as the trial cladding materials. The welding techniques and procedures were reported and discussed. Totally three welding techniques were adopted in the development of moderator cladding, and they are electron beam welding (EBW), laser welding (LW), and gas tungsten arc welding (GTAW). After welding, all cladded crucibles were evaluated with two kinds of leak testing, the helium leak test for minor leakage and the bubble test for major leakage. Due to the concern of the poor neutronics performance, Mo was not ideal for future moderator cladding application. Therefore, detailed characterization of the Mo cladded zirconium hydride was not pursued. Instead, the characterization of FeCrAl cladded zirconium hydride was performed. We discussed two material conditions (i.e., as machined and pre-oxidized conditions) prior to the cladding process, evaluated thermal stability of cladded zirconium hydride moderator through directly measuring hydrogen release, and characterized the zirconium hydride following the thermal desorption measurement. The results showed pre-oxidized FeCrAl is capable of efficiently preventing hydrogen release from the moderator assembly.

36 MATERIALS SCIENCE↗
DOE OSTI2016

A hydride reorientation model for irradiated zirconium alloy cladding

This paper presents a hydride reorientation model for irradiated, but de-fueled, zirconium-based alloy cladding tubes for water cooled reactors subjected to temperature and stress conditions of interest to storage of commercial spent nuclear fuel under dry, inert conditions. The present development builds upon and updates an existing hydride precipitation model developed in 2003, and was motivated by recent findings that the old model underestimated the amount of radial hydrides for applied hoop stresses below 100 MPa. The data used to support the development of the new model were obtained from an NFIR-sponsored experimental program on de-fueled cladding tubes irradiated in light-water reactors. The irradiated material consisted of three types of zirconium-based alloys sampled from high burnup fuel rods: cold-worked stress-relieved (CWSR) Zircaloy-4 (Zr-4), CWSR Zirlo{sup TM}, and recrystallized annealed (RXA) Zircaloy-2 (Zr-2). The majority of the hydride re-orientation tests for the irradiated cladding specimens were conducted under constant pressure with a mean cooling rate of about 1 Celsius degree/mn. The remaining tests were conducted at slower cooling rates with the pressure being kept proportional to the temperature change during the cooling process to simulate dry storage conditions. The parameters of the hydride re-orientation model were fitted to the test data. Fitting parameters were determined for (CWSR) Zircaloy-4 (Zr-4) and CWSR Zirlo{sup TM} respectively. The cooling rate effect was empirically accounted for in the model. General good agreement between model prediction and measured radial hydride content is achieved, with a standard error less than 12 ppm. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Post-quench ductility of zirconium alloy cladding materials - 2016-0088

The ductility of zirconium-based alloys after steam oxidation at high temperatures, followed by quenching in water, is of significant interest for the nuclear industry. High-temperature steam oxidation and quenching of nuclear fuel cladding occurs during postulated loss-of-coolant accidents. The industry is currently establishing in-reactor operation limits for peak cladding temperature and time at temperature based on the measured ductile-to-brittle transition of different cladding materials, both in the as-fabricated condition and using hydrogen charging as a surrogate for in-reactor operation. Aiming to study possible differences in post-quench ductility caused by differences in chemical composition, extensive testing on Zircaloy-4, ZIRLO{sup R}, and Optimized ZIRLO{sup TM} cladding materials as a function of hydrogen content, extent of oxidation, and peak cladding temperature was performed. The results showed similar ductile-to-brittle transition for the alloys studied despite differences in chemical composition, with a decrease in the extent of oxidation to reach the ductile to brittle transition with increasing hydrogen content. Microstructural characterization was performed to investigate the evolution of the oxygen-enriched layer that forms underneath the oxide, both during the time at high temperatures and during cooling. Significantly different morphologies of the oxygen-enriched layer were observed with different cooling rates, particularly for alloys containing niobium. Possible mechanisms for the development of the oxygen-enriched layer are discussed. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2024

Investigation of Degradation Mechanisms of Cr-coated Zirconium Alloy Cladding in Reactivity-Initiated Accident

Reactivity-initiated accident testing has been performed on two baseline fuel rodlets which were fabricated from Zr-4 cladding and utilized fresh UO 2 fuel pellets. These experiments are part of a larger series of transient irradiations to explore Cr-coated zirconium alloy cladding under Reactivity-initiated accident conditions to evaluate ballooning and burst and sever oxidation characteristic of late-stage failure. The first two irradiations have been completed and post-transient examinations have been performed. In one rod, ballooning and burst occurred, along with oxidation. The second rod leaked and as a result did not balloon as much during the transient. This second rod also experienced increased oxidation compared to the first due to higher temperatures.

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↗