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At least 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Breakaway oxidation of ZIRLO and optimized ZIRLO cladding per 10CFR50.46c criteria

The United States Nuclear Regulatory Commission (NRC) in 10CFR50.46c proposed changes to the loss of coolant accident (LOCA) criteria for LWR licensees including a new requirement to test cladding materials for breakaway oxidation. Cladding materials must be shown to resist breakaway oxidation long enough for the LOCA transient to be terminated. The onset of breakaway oxidation is characterized by a rapidly increasing rate of oxidation and the absorption of hydrogen resulting in a loss of cladding integrity. The proposed rule changes effectively require some manner of periodic and on-going testing to monitor the population of cladding for breakaway resistance. More than 200 lots of production cladding material have been tested in accordance with the draft regulatory guidelines. Testing has been conducted in a steam environment at the critical temperature for breakaway oxidation for durations that exceed the analytical endurance limit required to comply with the breakaway oxidation test criteria. The alloys that have been tested include both ZIRLO cladding and Optimized ZIRLO cladding. Breakaway oxidation testing consisted of three basic steps: preparation of test specimens, high temperature steam oxidation and post-test specimen evaluation. The results show there is a high probability that all ZIRLO and Optimized ZIRLO lots will pass production certification testing. Variations in response to breakaway oxidation testing were observed only in testing that ran well beyond the maximum LOCA duration of 2000 seconds. No indications of early breakaway oxidation were observed in the testing of production material that has been completed thus far. This provides confidence that a sampling approach to breakaway oxidation testing and monitoring remains a sound approach.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Evaluation of Sb-Nd and Te-Nd phases within the U-Zr fuel matrix and their interactions with HT9 alloy

Antimony (Sb) and tellurium (Te) were investigated as potential additives for U-10Zr (wt.%) metallic fuel to limit the fuel-cladding chemical interaction (FCCI) with HT-9 alloy. Neodymium (Nd) was utilized to simulate the formation of lanthanide-based solid fission products which are known to play a detrimental role in FCCI. Fuel alloys of U-Zr-Sb-Nd and U-Zr-Te-Nd were evaluated in their annealed condition and compared against their as-cast conditions. Isothermal diffusion couple experiments were performed between U-Zr-Nd, U-Zr-Sb-Nd, and U-Zr-Te-Nd against the cladding alloy HT9 to evaluate the effectiveness of the additives to stabilize Nd within the fuel alloys, as well as investigate the interaction regions that form between the different fuel alloys and HT9. Further, SbNd and Sb 3 Nd 4 , and TeNd are found to be the primary neodymium-based phases formed in the U-Zr-Sb-Nd and U-Zr-Te-Nd alloys, respectively. The zirconium-based phase, Zr 2 Sb, is also found to form within the former alloy. All phases were found to remain stable through the diffusion experiments and exhibited no interaction with HT9 constituent elements. Preferential interaction between Nd with additivities Te and Sb compared to constituting elements in HT9 was further verified based on density functional theory (DFT) calculated enthalpy of mixing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Rod Internal Pressure Distribution and Uncertainty Analysis Using FRAPCON

The discharge rod internal pressure (RIP) and cladding hoop stress (CHS) distributions are quantified for Watts Bar Nuclear Unit 1 (WBN1) fuel rods by modeling core cycle design data, operation data, and as-built fuel enrichments and densities of each fuel rod in FRAPCON-3.5. A methodology is developed that tracks intercycle assembly movements and assembly batch fabrication information to build individual FRAPCON inputs for each evaluated WBN1 fuel rod. An alternate model for the amount of helium released from the zirconium diboride (ZrB 2 ) integral fuel burnable absorber (IFBA) layer is derived and applied to FRAPCON output data to quantify the RIP and CHS for these types of fuel rods. SCALE/Polaris is used to quantify fuel rod–specific spectral quantities and the amount of gaseous fission products produced in the fuel for use in FRAPCON inputs. Fuel rods with ZrB 2 IFBA layers (i.e., IFBA rods) are determined to have RIP predictions that are elevated when compared to fuel rods without IFBA layers (i.e., standard rods) despite the fact that IFBA rods often have reduced fill pressures and annular fuel pellets. The primary contributor to elevated RIP predictions at burnups less than and greater than 30 GWd/tonne U is determined to be the total fuel rod void volume and the amount of released fission gas in the fuel rod, respectively. Cumulative distribution functions (CDFs) are prepared from the distribution of RIP and CHS predictions for all standard and IFBA rods. The provided CDFs allow for the determination of the portion of WBN1 fuel rods that exceeds a specified RIP or CHS limit. Results are separated into IFBA and standard rods so that the two groups may be analyzed individually. FRAPCON results are provided in sufficient detail to enable the recalculation of the RIP while considering any desired plenum gas temperature, total void volume, or total amount of gas present in the void volume. A method to predict the CHS from a determined or assumed RIP is also proposed that is based on the approximately linear relationship between the CHS and the RIP. Finally, improvements to the computational methodology of FRAPCON are proposed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Homogeneous hydride formation path in α-Zr: Molecular dynamics simulations with the charge-optimized many-body potential

A formation path for homogeneous γ hydride formation in hcp α-Zr, from solid solution to the ζ and then the γ hydride, was demonstrated using molecular static calculations and molecular dynamic simulations with the charge-optimized many-body (COMB) potential. Hydrogen has limited solubility in α-Zr. Once the solubility limit is exceeded, the stability of solid solution gives way to that of coherent hydride phases such as the ζ hydride by planar precipitation of hydrogen. At finite temperatures, the ζ hydride goes through a partial hcp-fcc transformation via 1/3 <1¯100> slip on the basal plane, and transforms into a mixture of γ hydride and α-Zr. In the ζ hydride, slip on the basal plane is favored thermodynamically with negligible barrier, and is therefore feasible at finite temperatures without mechanical loading. The transformation process involves slips of three equivalent shear partials, in contrast to that proposed in the literature where only a single shear partial was involved. The adoption of multiple slip partials minimizes the macroscopic shape change of embedded hydride clusters and the shear strain accumulation in the matrix, and thus reduces the overall barrier needed for homogeneous γ hydride formation. In conclusion, this formation path requires finite temperatures for hydrogen diffusion without mechanical loading. Therefore, it should be effective at the cladding operating conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Modeling of hydride precipitation and re-orientation

In this report, we present a thermodynamic-­based model of hydride precipitation in Zr-based claddings. The model considers the state of the cladding immediately following drying, after removal from cooling-pools, and presents the evolution of precipitate formation upon cooling as follows: The pilgering process used to form Zr-based cladding imparts strong crystallographic and grain shape texture, with the basal plane of the hexagonal α-Zr grains being strongly aligned in the rolling-­direction and the grains are elongated with grain size being approximately twice as long parallel to the rolling direction, which is also the long axis of the tubular cladding, as it is in the orthogonal directions.

36 MATERIALS SCIENCE↗
DOE OSTI2015

Structure and mechanical properties of coatings fabricated by nonvacuum electron beam cladding of Ti-Ta-Zr powder mixtures

In this paper structural investigations and mechanical tests of Ti-Ta-Zr coatings obtained on surfaces of cp-titanium workpieces were carried out. It was found that the coatings had a dendrite structure; investigations at high-power magnifications revealed a platelet structure. An increase of tantalum concentration led to refinement of structural components. The microhardness level of all coatings, excepting a specimen with the maximum tantalum content, was 370 HV. The microhardness of this coating reached 400 HV. The ultimate tensile strength of cladded layers varied from 697 to 947 MPa. Adhesion tests showed that bimetallic composites were characterized by high bond strength of cladded layers to the substrate, which exceeded cp-titanium strength characteristics.

36 MATERIALS SCIENCE↗
DOE OSTI2025

DISSOLUTION OF SURROGATE U-Zr FUEL USING ALNIFLEX CONDITIONS

Non-aluminum clad spent nuclear fuel (NASNF) stored in L Basin at the Savannah River Site (SRS) is widely varied in fuel composition, design, packaging, and physical condition. The complexity of the NASNF inventory presents significant challenges, and technology development is necessary for successful disposition. One such fuel in the inventory is metallic uranium-zirconium (U-Zr) alloy fuel, the focus of this study. Electrolytic or nitric acid only dissolution of metallic U-Zr alloy can form insoluble zirconium oxide, which results in up to 52% loss of U to insoluble solids, and can be subject to potentially uncontrolled oxidation reactions [1, 2]. The AlNiflex process was determined to be a viable dissolution flowsheet for the U-Zr alloy fuel. Under a narrow set of solution concentrations, a combination of hydrofluoric acid (HF), nitric acid (HNO3), aluminum nitrate (Al(NO3)3), and hexavalent chromium can safely dissolve U-Zr intermetallic alloys, keep Zr soluble, and not significantly corrode stainless steel (SS) vessels [3, 4.

Gogolski, Jarrod M. [Savannah River National Labor↗
DOE OSTI2016

Development of FeCrAl-ODS steels for ATF cladding

As an enhanced accident tolerant fuel (ATF) for LWR at severe accident condition, FeCrAl-ODS ferritic steels have been developed, and their applicability is being studied under the Japanese MEXT program. Alloying design of FeCrAl-ODS ferritic steels has been conducted, focusing on effect of mutual interaction of Cr-Al on oxidation resistance and α'-embrittlement. Effects of co-addition of Zr and excessive oxygen have also been studied to improve high-temperature strength. The ring tensile test of the manufactured claddings showed that tensile strength increases by Zr-addition and Zr-added FeCrAl-ODS steels have the highest strength among the existed steels at entire temperature range from room temperature up to 1000 Celsius degrees. Although Zr-addition sometimes leads to growth of alumina scale by rapid inward diffusion of oxygen through zirconia inside alumina, oxidation rate of Zr-added FeCrAl-ODS steels was successfully reduced to the level of Zr-free specimen by controlling oxygen content to form Y{sub 2}Zr{sub 2}O{sub 7} inside ODS steel matrix. This finding makes possible to improve not only high-temperature strength but also steam oxidation resistance in the Zr-added FeCrAl-ODS steels. (authors)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Advantages gained by Optimized ZIRLO and AXIOM PWR cladding materials

In support of industry's demands for higher performance cladding alloy materials, the Optimized ZIRLO alloy was developed as an evolutionary improvement of the ZIRLO alloy to obtain a significant improvement in corrosion performance and hydrogen pickup while maintaining the overall robust performance of ZIRLO alloy. Currently with operating experience in over 40 plants and burnup reaching over 70 GWd/MTU, a majority of the Westinghouse-fueled US PWR fleet, covering almost all Westinghouse fuel designs, has adopted full regions of Optimized ZIRLO cladding. Across all of these applications, Optimized ZIRLO cladding has demonstrated greatly improved corrosion performance and structural stability. These demonstrated improvements offer increased operational margin and flexibility, in support of industry trends toward increased core operating fuel duties and evolving regulatory requirements such as accident scenarios. Underpinning these benefits are the extensive post irradiation exam (PIE) data obtained from both pool-side and hot-cell exams for Optimized ZIRLO cladding. Building on the successes of Optimized ZIRLO cladding, and in recognition of increasingly challenging fuel management practices, and revised LOCA regulatory rules restricting cladding hydrogen content, AXIOM cladding is being rolled out for full lead test assembly irradiation. The AXIOM alloy, Westinghouse's next generation robust alloy targeting very high fuel duties, is designed to provide improved corrosion resistance, lower hydrogen pick-up, and lower creep and growth compared to current Westinghouse products. AXIOM cladding has been irradiated in a variety of reactors worldwide, including both commercial and test reactors. Extensive PIE data is available for various burnup levels, with leading rods approaching 75 GWd/MTU. The AXIOM composition was selected to ensure enhanced operating performance on multiple fronts, including the optimum combination of corrosion resistance and wear resistance, reduced hydrogen pick-up, improved dimensional stability, and robustness in aggressive coolant chemistry conditions. The present paper provides high burnup performance data, capturing the advantages gained by introducing Optimized ZIRLO and AXIOM alloy as PWR cladding materials. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

A 100 MWe advanced sodium-cooled fast reactor core concept

An Advanced sodium-cooled Fast Reactor core concept (AFR-100) was developed targeting a small electrical grid to be transportable to the plant site and operable for a long time without frequent refueling. The reactor power rating was strategically decided to be 100 MWe, and the core barrel diameter was limited to 3.0 m for transportability. The design parameters were determined by relaxing the peak fast fluence limit and bulk coolant outlet temperature to beyond irradiation experience assuming that advanced cladding and structural materials developed under US-DOE programs would be available when the AFR-100 is deployed. With a de-rated power density and U-Zr binary metallic fuel, the AFR-100 can maintain criticality for 30 years without refueling. The average discharge burnup of 101 MWd/kg is comparable to conventional design values, but the peak discharge fast fluence of {approx}6x10{sup 23} neutrons/cm{sup 2} is beyond the current irradiation experiences with HT-9 cladding. The evaluated reactivity coefficients provide sufficient negative feedbacks and the reactivity control systems provide sufficient shutdown margins. The integral reactivity parameters obtained from quasi-static reactivity balance analysis indicate that the AFR-100 meets the sufficient conditions for acceptable asymptotic core outlet temperature following postulated unprotected accidents. Additionally, the AFR-100 has sufficient thermal margins by grouping the fuel assemblies into eight orifice zones. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Fuel performance analysis of FeCrAl cladding during LWR operation

This work focuses on modeling the integral thermo-mechanical performance of FeCrAl cladded fuel during normal operation in light-water reactors. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON. These simulations identify the effects of the mechanical-stress and irradiation responses of FeCrAl and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (∼4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

A Multiphysics Evaluation of Annular Uranium-Zirconium Metallic Fuels [Poster]

This study examines the performance of U-10Zr annular metallic fuel rodlets which were experimentally evaluated as part of the Advanced Fuels Campaign (AFC). The AFC mission is to develop novel fuel technologies and facilitate the implementation of those technologies by industry partners. A key objective is to improve steady-state and transient performance over current fuel types. The experiments of interest in this study included annular metallic U-Zr fuel rodlets within HT-9 cladding which were placed in SS-316 capsules and inserted in the Advanced Test Reactor (ATR). Certain mechanical and thermal conditions cannot be directly evaluated through experiments and fuel performance modeling is used to shed light on this evolution over time. In this study, BISON Multiphysics simulations are leveraged to investigate the state of the fuel system throughout and after the experimental conditions.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Fuel cladding post-quench LOCA embrittlement: mechanical test relevance

During a loss-of-coolant-accident (LOCA), the fuel cladding is submitted to a high temperature increase in a steam environment, along with an internal pressure increase which would eventually lead to ballooning and burst of the cladding. Steam could then penetrate through the burst opening and lead to inner surface oxidation and secondary-hydriding of the cladding. Post-quench cladding rupture can then occur either in two-sided oxidized area with wall thinning due to ballooning, or in single-sided oxidized area with large hydrogen content. Common tests involved in the evaluation of the post-quench ductility are Ring Compression Tests (RCT), Expansion Due to Compression Test (EDC) or Axial Tensile Tests (AT). Questions arise about these mechanical tests, as for instance, the influence on sample failure of the oxide layer thickness partitioning between inner and outer diameter. A systematic study was carried out to compare several mechanical tests on both single-sided and two-sided Zircaloy-4 oxidized samples. Five different mechanical tests were compared, including RCT: C-shaped ring Compression Test (CCT), Axial Tensile test (AT), Ring Tensile Test with well-defined machined gage section (RTT) and plain Ring Tensile Test (pRTT). All samples were taken from cladding tubes which had been oxidized in steam at 1200 Celsius degrees and then water-quenched. Mechanical tests were performed at room temperature and two to three samples were tested for each oxidation conditions and for every sample geometry to assess the reproducibility. Metallographic examinations were carried out to measure the ZrO{sub 2} and αZr(O) layer thicknesses and the ex-β/α(O) phase fraction. Mean hydrogen content measurements were also performed systematically on each sample. Results indicate that the different mechanical tests have not the same sensitivity to the oxide layer partitioning: RTT and pRTT load at failure seems to depend mainly on the cumulated oxide layer thickness rather than on its partitioning, whereas this parameter has an impact on CCT and AT test results. For all tensile tests, stresses at failure were derived from load-displacement curves. The failure of brittle samples was assumed to be governed by linear elastic fracture mechanics. Using conventional stress intensity factor solution, a value of the material fracture toughness controlling the sample failure was determined from experimental results and discussed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Optimizing Hydride Stability in U-ZrH x Nuclear Fuel: The “Goldilocks Radius”

Nuclear-powered microreactors show great promise for opening new nuclear energy markets due to the flexibility offered by their rapid/streamlined in-factory fabrication, transportability, and self-regulating nature. The economic benefits of any commercialized nuclear reactor, however, rely on the system’s ability to produce large amounts of heat and efficiently convert that heat into electrical power reliably for long periods of time. Uranium-zirconium hydride (U-ZrH x ) is currently being considered for compact reactor designs because it is a well-known nuclear fuel system that is self-moderating, but this fuel, which has historically been used for research reactors, has not been optimized for commercial power production. Here, this paper analyzes the hydride stability of standard 304 stainless steel–clad U-ZrH x fuel under commercially relevant conditions. Fuel element design parameters, including physical dimensions, as-fabricated hydrogen content, burnup, peak fuel temperature, temperature gradient, operational fuel cycle duration, and volumetric heat generation rate, are discussed with a focus on hydrogen distribution and phase stability within the fuel element. Hydride stability declines more rapidly as the coolant temperature, burnup, and fuel cycle duration increase. Using a fuel-cladding gap material with heat transfer properties superior to air, such as helium or sodium, is essential to prolonging fuel hydride stability. The fuel’s physical dimensions are also important. At very small fuel diameters, the H/Zr ratio in the fuel meat decreases too rapidly due to the hydrogen content’s dependence on fuel meat volume. Conversely, the fuel meat temperature and temperature gradient exacerbate hydrogen loss at very large fuel diameters. We find that the most important parameter to consider when optimizing the hydride stability of U-ZrH x fuel is the relationship between the fuel meat radius and the power density in the fuel. A simple equation is empirically determined that relates the “Goldilocks radius,” that is, the fuel radius for which the H/Zr ratio is most stable, to the power density in the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Nuclear fuel behavior at high temperature: new insights from post-test examinations on the VERDON-1 sample

In order to reduce the uncertainties concerning Source Term assessment under Light Water Reactor (LWR) severe accident conditions, 4 VERDON tests were recently performed. They have been defined in the framework of the International Source Term Program and carried out at CEA Cadarache, in the LECA-STAR facility. Among these 4 VERDON tests, VERDON-1 test was devoted to high burn-up UO{sub 2} fuel behavior and FP release under H{sub 2} reducing conditions at very high temperature. This test was performed with the Release Circuit, which is devoted to the precise characterization of the FP release, in terms of FP release kinetics and fission products (FPs) released fraction. The main components of the circuits are a furnace heating the fuel sample, a filter trapping the emitted aerosols and gamma spectrometry detectors for online and post-test measurements. VERDON-1 test was conducted up to about 2600 Celsius degrees without reaching global fuel relocation, allowing thereby post-test micro-analyses. After a brief description of the fuel sample nature and the experimental set-up, we present the progress of the experimental sequence and the associated gamma spectrometry measurements. We then focus on the major results of the VERDON-1 tests regarding the fuel global behavior. These results are obtained by two complementary techniques. First, the fuel sight gamma station can be used either on-line to detect fuel relocation or just after the test, as a longitudinal bench, to get preliminary information on the global shape of the fuel sample inside the furnace. Secondly, these information are completed by post-test examinations performed on the fuel sample after extraction from the furnace. Complete qualitative and quantitative characterizations, including EPMA (Electron Probe Micro-Analysis) and SEM (Scanning Electron Microscopy) analysis, were performed for the first time at the CEA Cadarache on a sample that has been submitted to such severe accident conditions. These results show an advanced state of sample degradation, with locally previously molten areas, meaning that fuel global relocation was imminent by the end of thermal-hydraulic sequence. The relocation temperature is compared to that of previous VERCORS tests and discussed regarding the atmosphere of the test and possible UO{sub 2}-ZrO{sub 2} interaction. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Modeling of High-Temperature Corrosion of Zirconium Alloys Using the eXtended Finite Element Method (X-FEM)

Oxidation modeling in modern nuclear fuel performance codes is currently limited by the lack of coupling with mechanics, thus preventing proper description of how high-temperature oxidation impacts mechanical properties. This is mostly due to the fact that the finite difference formalism adopted in corrosion models is incompatible with the direct coupling with mechanics in the finite element modeling employed in modern nuclear fuel performance codes. In this study, a physically based zirconium alloy corrosion model called the Coupled-Current Charge Compensation (C4) model, which was initially developed for operating temperature conditions, has been updated to include high-temperature corrosion in order to provide additional critical information (e.g., oxygen concentration profile) under loss-of-coolant accident (LOCA) conditions—information lacking in existing empirical models. The C4 model was implemented in the MOOSE finite-element framework developed at Idaho National Laboratory, enabling it to be used in the BISON nuclear fuel performance code based on the MOOSE framework. To precisely track the different interfaces at a relatively low computational cost, the eXtended Finite Element Method (X-FEM) was applied in MOOSE. The model’s results were compared to those of existing empirical models as well as metallographic analysis of high-temperature oxidized Zircaloy-4 coupons. Oxygen diffusivities in the a and ß phases resulting from this comparison closely agree with those found in the literature. The C4 model implemented with X-FEM in MOOSE now has the capability to accurately predict oxide, oxygen-stabilized a, and prior ß phase layer growth kinetics under isothermal exposure at high temperature (1000–1500°C). Furthermore, in contrast with the empirical models, the C4 model accounts for the finite thickness of the fuel cladding. It can predict the oxygen concentration profile evolution through the whole cladding, enabling evaluation of the remaining ductile thickness—a crucial variable for modeling the mechanical behavior of the fuel cladding under LOCA. Furthermore, this implementation allows direct coupling with mechanics, at a low computing cost, using finite-element-based nuclear fuel performance codes such as BISON.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Oxidation behavior of Zry-4 in steam-air mixtures at high temperature

This study dealt with the oxidation behavior of Zry-4 nuclear fuel cladding under a severe nuclear reactor accident scenario occurring in a light-water reactor. Influence of the reaction extent on cladding mechanical properties was also treated. Zry-4 fuel cladding segments were exposed to steam atmosphere containing different amount of air in composition ranging from 0 to 100 vol. %. Exposure times covered both pre- and post-breakaway regimes. Metallographic observations along with microhardness testing were carried out to study the effect of air fraction on the material microstructure. Hydrogen pick-up was measured and residual mechanical properties were assessed employing Ring Compression Tests. The results show that the addition of air in steam can lead to more than 300 % increase of weight gain in case of ∼50 % or higher air fraction after the exposure of 60 min., namely, 'late post-breakaway regime'. Approximately three times thicker oxide scales were observed in such cases, indicating that most of the weight gain was due to the oxide layer increase. Nitride phase was preferentially observed close to the interface between oxide and metal layers. The α-Zr(O) sublayer thickness decreases with the increasing fraction of air in steam. Microhardness, measured within the prior β-phase region, slightly depends on the air fraction. Hydrogen uptake, on the other hand, depends strongly on the air fraction in steam. In the case of shorter exposure times, the effect of air addition in steam was suppressed. Eventually, specimen plastic strains drastically reduced in the late post-breakaway regime. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗