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At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
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 OSTI2016

Post quench ductility of Zirconium alloy cladding materials - 6.2 STP-2016-0088

The ductility of zirconium-based alloys after steam oxidation at high temperature, 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 (LOCA). 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 (PQD) caused by differences in chemical composition, extensive testing has been performed 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. The results show similar ductile-to-brittle transition for the alloys studied, in spite of the 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 appears underneath the oxide, both during the time at high temperature and during cooling. Significantly different morphologies of the oxygen enriched layer are 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 OSTI2016

Investigations into the Gas Phase Reactions of Zirconium Tetrachloride (ZrCl) from UNF Cladding

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. A step in a strategy to recycle zirconium from used cladding materials is the high yield chlorination of Zircaloys to produce zirconium tetrachloride (ZrCl{sub 4}). The ZrCl, is contaminated with other metal chloride species that originate from the alloying metals as well as contaminants from oxide layers on the cladding, trace fission products, fuel residues, activation byproducts, and tritium. The properties of these complex mixtures must be better understood before purification protocols can be developed that will allow zirconium chloride to be recycled further. First order thermodynamic modeling of high temperature mixtures of ZrCl, and other metal chlorides in the gas phase may not accurately predict their properties. Computational modeling studies indicate that mixed metal adducts may form under the conditions of sublimation-based purification protocols. In light of these predictions, alternate strategies may also be needed to obtain pure ZrCl, from such mixtures. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

AREVA enhanced accident tolerant fuel program - Current results and future plans

The purpose of this paper is to provide a brief overview of the enhanced accident tolerant fuel (EATF) concepts investigated by the AREVA team in Phase 1 to support the U.S Department of Energy Enhanced Accident Tolerant Fuel program and concerning light water reactors. The main goal of the 'fuel pellets with additive' concept is to improve the economics, performance and operational flexibility of fuel in light water reactors while featuring improved safety performance under normal and off-normal conditions. Within this concept 2 nuclear fuels were explored: Cr{sub 2}O{sub 3} doped UO{sub 2} pellet and composite UO{sub 2} pellets. The second concept is the coating of zirconium cladding whose main goal is to reduce cladding oxidation and hydrogen generation during accident conditions. The coating materials were chosen to be compatible with the specific deposition process, to provide robustness in the nuclear environment and have a relatively small neutron cross section. Three coatings were explored: chromium coated claddings, electrophoretic deposition of coating, and MAX phase coatings. MAX phases are a family of the layered compounds with a general formula M{sub n+1}AX{sub n} with n = 1, 2, or 3, where M is a transition metal, A is an A-group element, and X is either C or N. Another improvement to the fuel cladding that was considered is to replace the zirconium cladding altogether with an alternative cladding material which can substantially improve reactor response during postulated accident conditions. Two concepts were investigated: molybdenum claddings and SiC/SiC composite cladding.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Corrosion resistant coatings for zirconium-alloy cladding with improved accident tolerance

The Westinghouse-led accident tolerant fuel (ATF) program is evaluating Accident Tolerant Fuel (ATF) coatings of MAX phase (Ti{sub 2}AlC) and multi-layer Ti{sub x}Al{sub 1-x}N/TiN to enhance the oxidation resistance of zirconium alloy fuel cladding in both normal operation and beyond design-based conditions of light water reactors. A variety of coating technologies have been applied to deposit these coating materials onto Zircaloy-4 or Zirlo sheets and tubes, including cold spray and cathodic arc-physical vapor deposition (CA-PVD). Neutronic analysis using the elemental compositions of the coating materials showed that the coating thicknesses should be less than 30 μm for most coating materials to avoid non-negligible economic penalty. The coated samples were autoclave tested in water at 360 Celsius degrees and 150 bar or in steam at 427 Celsius and 103 bar for corrosion resistance evaluation. The initial Ti{sub 2}AlC coating made by using large particle size MAX phase powder and cold spray deposition did not provide the desired corrosion resistance due to un-optimized microstructure and porosity. Subsequent optimization including reduction in particle size and tuning of cold spray deposition parameters resulted in denser coatings less than 30 microns in thickness. Optimization of process conditions resulted in reduced oxidation kinetics and showed that the coating acts as an effective barrier to prevent the zirconium substrate from oxidizing in accident condition temperatures. High purity Ti{sub 2}AlC powder is needed to further improve oxidation resistance of the Ti{sub 2}AlC coating. The oxidation resistance of pure monolithic MAX phase compound was also found to be remarkable, and the weight gain was only 1.85 mg/cm{sup 2} for 96 hours exposure in 1200 C. degrees steam, which is consistent with the literature. The corrosion resistance of the TiAlN/TiN multilayer coating was excellent. The weight gain was negligible and no oxidation was observed on the surface of zirconium-alloy at the coating-substrate interface. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Sensitivity analysis for characterizing the impact of HNGD model on the prediction of hydrogen redistribution in Zircaloy cladding using BISON code

Hydrogen in zirconium cladding is able to precipitate into zirconium hydrides which impacts cladding integrity. The Hydride Nucleation-Growth-Dissolution (HNGD) model in the BISON code accounts for the precipitation and dissolution kinetics of hydride in Zircaloy material. This paper presents global sensitivity analyses of the HNGD model aiming to enhance our understanding of the hydride precipitation phenomena by quantifying the variance that key parameters have on the prediction of hydrogen behavior under various environmental conditions. Here, model predictions are compared to experimental data obtained under two different conditions: 1) with uniformly precharged specimens subjected to a linear thermal gradient, and 2) specimens precharged with a cathodically applied hydride rim at one end of the sample and subjected to an asymmetric thermal gradient. The Sobol sensitivity analysis identifies the key parameters in the HNGD model for both types of specimens. For linear temperature cases, the heat of transport dominates the accuracy of predictions when no precipitation occurs at the cold end, while Terminal Solid Solubility for Dissolution (TSSD) is the most important parameter when precipitation occurs. A large variation in the predicted hydrogen concentration profiles is found in the range of high TSSD due to the occurrence of precipitation. For asymmetric temperature cases, the solubility coefficient gives the largest impact on the predicted hydrogen distribution, as it determines the amount of solute hydrogen dissolved from the initially applied hydride rim. A large discrepancy in hydrogen distribution between simulations and experiments exists with the asymmetric specimens because BISON simulations fail to predict the precipitation of hydride at the cooler end. Comparative studies using former and updated models verifies the significant impact of the hydride growth mechanism on predicted hydrogen concentration profiles. In particular, when hydride initially exists, changes in TSSD generate a large variation in the predicted amount of precipitation by hydride growth, giving large uncertainty in predicting the hydrogen distribution over the sample length. The outputs characterize the significant impact of the hydride growth mechanism in the HNGD model on predicting hydrogen behavior, and improve the understanding of the precipitation of hydride in Zircaloy cladding within a range of expected environmental conditions. The analyses indicate work is still needed to improve the hydride solvus models in the BISON code to accurately predict experimentally observed hydride concentrations and distributions.

36 MATERIALS SCIENCE↗
DOE OSTI2024

Bench-Scale Electrolytic Dissolution of Quarter-Scale FCA Cans

In 2016, the Savannah River National Laboratory (SRNL) led, in support of and under sponsorship of the Department of Energy’s National Nuclear Security Administration (DOE/NNSA) Office of Material Management and Minimization (M3), the removal and transfer of the plutonium based Fast Critical Assembly (FCA) fuel from the Japan Atomic Energy Agency (JAEA) Tokai facility to the Savannah River Site (SRS). The team also included JAEA, multiple organizations in Savannah River Nuclear Solutions (SRNS), International Nuclear Services, and many other entities. The FCA fuel removal project completion was a key deliverable for M3 to the 2016 Nuclear Security Summit and constituted the largest inventory of weapons-usable plutonium removed under the nonproliferation program. The FCA materials consist of thousands of stainless steel (SS) clad plates and hundreds of SS clad rods. The FCA fuel elements were packaged in a carrier can and stored at SRS pending disposition of the fuel. Following an assessment of candidate disposition options, SRNS identified electrolytic dissolution (ED) as the most promising disposition option for the FCA plates and their preferred option was endorsed by DOE. This option entails electrochemically dissolving the entire FCA carrier can with fuel elements and was based on bench-scale laboratory testing and historical work on processing SS-clad and zirconium-clad uranium-based fuel in the H-Canyon electrolytic dissolver (last operated in 1980). The FCA plate consists of a plutonium-aluminum metal alloy core hermitically sealed in SS cladding.

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

Dynamic Probabilistic Risk Assessment Based Response Surface Approach for FLEX and Accident Tolerant Fuels for Medium Break LOCA Spectrum

After the Fukushima Daiichi Accident, the safety features such as accident tolerant fuel (ATF) and diverse and flexible coping strategies (FLEX) for existing nuclear fleets are being investigated by the US Department of Energy under the Light Water Reactor Sustainability Program. This research is being conducted to quantify the risk-benefit of these safety features. Dynamic probabilistic risk assessment (DPRA)-based response-surface approach has been presented to quantify the FLEX and ATF benefits by estimating the risk associated with each option. ATFs with multilayered silicon carbide (SiC), iron-chromium-aluminum, and chromium-coated zirconium cladding were considered in this study. While these ATF candidates perform better than the current zirconium cladding (Zr), they may introduce additional failure modes in some operating conditions. The fuel failure analysis modules (FAMs) were developed to investigate ATF performance. The dynamic risk assessments were performed using RAVEN, a DPRA tool, coupled with RELAP5 and FAMs. A cumulative distribution function-based index provided a mean of comparing the benefits of safety enhancements. For medium break loss of coolant accidents, FLEX operational timing window for each fuel type was estimated. Among these ATF candidates, SiC-type ATF was the most beneficial candidate for an increased safety margin than Zr-based fuel and was found to complement FLEX strategies in terms of risk and coping time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Understanding corrosion and hydrogen pickup of zirconium fuel cladding alloys: The role of oxide microstructure, porosity, sub-oxides, and second-phase particles - 2016-0071

We used a range of advanced microscopy techniques to study the microstructure, nanoscale chemistry, and porosity in zirconium alloys at different stages of oxidation. Samples from both autoclave and in-reactor conditions were available, including ZIRLO{sup TM}, Zr-1.0Nb, and Zr-2.5Nb samples with different heat treatments. Scanning transmission electron microscopy (STEM), transmission Kikuchi diffraction (TKD), and automated crystal orientation mapping with TEM were used to study the grain structure and phase distribution. Significant differences in grain morphology were observed between samples oxidized in the autoclave and in-reactor, with shorter, less well-aligned monoclinic grains and more tetragonal grains in the neutron-irradiated samples. A combination of energy-dispersive X-ray mapping in STEM and atom probe tomography analysis of second-phase particles (SPPs) can reveal the main and minor element distributions respectively. Neutron irradiation seems to have little effect on promoting fast oxidation or dissolution of β-niobium precipitates but encourages the dissolution of iron from Laves-phase precipitates. An electron energy-loss spectroscopy (EELS) analysis of the oxidation state of niobium in β-niobium SPPs in the oxide revealed the fully oxidized Nb{sup 5+} state in SPPs deep into the oxide but Nb{sup 2+} in crystalline SPPs near the metal-oxide interface. EELS analysis and automated crystal orientation mapping with TEM revealed Widmanstatten-type suboxide layers in some samples with the hexagonal ZrO structure predicted by ab initio modeling. The combined thickness of the ZrO suboxide and oxygen-saturated layers at the metal-oxide interface correlated well to the instantaneous oxidation rate, suggesting that this oxygen-rich zone is part of the protective oxide that is rate limiting in the transport processes involved in oxidation. Porosity in the oxide had a major influence on the overall rate of oxidation, and there was more porosity in the rapidly oxidizing annealed Zr-1.0Nb alloy than in either the recrystallized alloy or the similar alloy exposed to neutron irradiation. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Behavior of fuel with zirconium alloy cladding in reactivity-initiated accident and loss-of-coolant accident - 2016-0090

Along with our colleagues at the Japan Atomic Energy Research Institute (JAERI) and Japan Atomic Energy Agency (JAEA), we have performed extensive research programs for more than 2 decades and have developed a better understanding of fuel behavior under accident conditions. Data and findings from these research programs have provided the technical basis directly and indirectly for regulatory criteria in Japan and other countries. This paper reviews and summarizes the major outcomes from the research programs performed at JAERI and JAEA facilities and identifies further research needs. The programs aim to generate data and models to (1) evaluate the adequacy of present safety criteria and safety margins; (2) create a database for the future regulation of higher-burnup UO{sub 2} and mixed-oxide (MOX) fuels, new cladding, and pellets; (3) provide adequately mechanistic computer codes for regulatory application; (4) promote a better understanding of phenomena that appear in high-burnup regions and MOX fuels, such as rim effect and an effect of plutonium agglomerates in MOX fuels, and to evaluate those effects on fuel behavior in accident conditions; and (5) to assess fuel behaviors with higher fuel duty, such as plant power uprates, longer operating cycles, and water chemistry changes. The programs comprise reactivity-initiated accident studies, including pulse-irradiation experiments in the Nuclear Safety Research Reactor, cladding mechanical tests, and the development and verification of RANNS; loss-of-coolant accident tests, including integral thermal shock tests, oxidation rate measurements, and cladding mechanical tests; development and verification of a computer code FEMAXI-6 that simulates high-burnup fuel behavior under normal operating and abnormal transient conditions; and studies on phenomena specific to high-burnup regions, including thermal property measurements, experimental and analytical simulation of rim structure formation, cladding mechanical tests on effects of hydrides, etc. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Silicon carbide reinforced zirconium based cladding

A method for making an improved nuclear fuel cladding tube includes reinforcing a Zr alloy tube by first winding or braiding ceramic yarn directly around the tube to form a ceramic covering, then physically bonding the ceramic covering to the tube by applying a first coating selected from the group consisting of Nb, Nb alloy, Nb oxide, Cr, Cr oxide, Cr alloy, or combinations thereof, by one of a thermal deposition process or a physical deposition process to provide structural support member for the Zr tube, and optionally applying a second coating and optionally applying a third coating by one of a thermal deposition process or a physical deposition process. If the tube softens at 800° C.-1000° C., the structural support tube will reinforce the Zr alloy tube against ballooning and bursting, thereby preventing the release of fission products to the reactor coolant.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Thermal Calculations to Determine Maximum Zircaloy Temperatures During Reactor Operations for Campaign 1 Fuel Assemblies

The H-Canyon Facility plans on processing bundles from Non-Aluminum Spent Nuclear Fuel (NASNF) Risk Group 1 (UO 2 core fuels). Uranium-zirconium (UZr) alloys and zirconium clad UO 2 fuel containing UZr intermetallic compound have the potential to cause explosions during dissolution. UZr intermetallic typically only forms when uranium and zirconium (or zirconium alloy) cladding are in contact at high temperatures. Thermal modeling was performed to determine the maximum cladding temperature for Campaign 1 Zircoloy cladded fuel assembles during reactor operation to help determine if the UZr intermetallic compound could have formed during reactor operations. The temperature of 350°C was chosen as a conservative minimum value for the possible formation of intermetallic UZr compound. The modeling approach, inputs, and results are outlined in this document.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Anisotropic Azimuthal Power and Temperature distribution on FuelRod: Impact on Hydride Distribution

The degradation of the zirconium cladding may limit nuclear fuel performance. In the high temperature environment of a reactor, the zirconium in the cladding corrodes, releasing hydrogen in the process. Some of this hydrogen is absorbed by the cladding in a highly inhomogeneous manner. The distribution of the absorbed hydrogen is extremely sensitive to temperature and stress concentration gradients. The absorbed hydrogen tends to concentrate near lower temperatures. This hydrogen absorption and hydride formation can cause cladding failure. This project set out to improve the hydrogen distribution prediction capabilities of the BISON fuel performance code. The project was split into two primary sections, first was the use of a high fidelity multi-physics coupling to accurately predict temperature gradients as a function of r, θ , and z, and the second was to use experimental data to create an analytical hydrogen precipitation model. The Penn State version of thermal hydraulics code COBRA-TF (CTF) was successfully coupled to the DeCART neutronics code. This coupled system was verified by testing and validated by comparison to FRAPCON data. The hydrogen diffusion and precipitation experiments successfully calculated the heat of transport and precipitation rate constant values to be used within the hydrogen model in BISON. These values can only be determined experimentally. These values were successfully implemented in precipitation, diffusion and dissolution kernels that were implemented in the BISON code. The coupled output was fed into BISON models and the hydrogen and hydride distributions behaved as expected. Simulations were conducted in the radial, axial and azimuthal directions to showcase the full capabilities of the hydrogen model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗