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At least 91 records · Page 5
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DOE OSTI2024

Impact of LWR assembly structural features on cladding burst behavior under LOCA conditions

This provides an initial scoping study on clad balloon and burst behavior for burnup extension of reactor fuel. The associated issues with burnup extension are fuel fragmentation, relocation, and dispersal in the event of cladding failure. The general finding of this work is that the structural features, spacer grids and mixing vanes, locally suppress cladding deformation but have little impact on the overall clad performance during loss-of-coolant accidents. The work detailed in a previous report by Capps et al. focused on core optimization via neutronics, thermal hydraulics and thermomechanical analysis for burnups beyond 62 GWD/tU and enrichments above 5%. Uncertainty of fuel fragmentation relocation and dispersal in high-burnup rods during accident conditions was also investigated. The dispersal aspect of fuel fragmentation depends on cladding rupture. Thus, assessing uncertainties in the rupture behavior is helpful in estimating the dispersal of high-burnup fuel. This study builds on the previous work by assessing the impact of assembly structural features on cladding balloon and burst behavior in a full-length fuel rod. In this work, the BISON fuel performance code was used to generate 2D radial and height meshes containing structural features commonly used in nuclear fuel assemblies. First, meshes were generated with spacer grids. Results were then compared to the cladding burst temperature and balloon strain results from the previous work. A mesh sensitivity study was performed to ensure that mixing vanes and spacer grid effects were appropriately considered, resulting in a more refined mesh than the previous study. The balloon deformation and burst times of the cladding were compared to the original case. Consideration was also given to the effect of rod initial pressure. In conclusion, 3D quarter rod simulations were also performed and found good agreement with the 2D simulations in clad deformation and reasonable agreement in burst times.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Thermal mechanical assessment of a SiC-SiC-composite clad fuel pin concept in a light water reactor environment

Accident Tolerant Fuels (ATFs) are designed to increase coping time following an accident scenario while preserving or improving current steady state reactor operational performance. A potential ATF concept is SiC-SiC composite claddings. Fuel performance simulations were conducted on a SiC-SiC based cladding concept utilizing a multilayered approach for improved performance. This cladding concept referred to as the Duplex concept is a duplex structure composed of a monolithic SiC layer placed on the outside of a SiC-SiC composite. A liquid metal is added to fuel-cladding gap for improved heat dissipation from the fuel. The monolithic SiC layer is used to improve the coolant corrosion characteristics and protect the SiC-SiC composite layer from exposure to the coolant. The fuel performance code BISON was used to conduct fuel performance simulations on the cladding concepts. Comparisons are made with a current prototypic fuel rod design (UO 2 fuel enclosed in Zircaloy-4 cladding). Representative steady-state cases were considered for normal power and two cycle power histories. Additionally, a PCI ramp case was simulated to analyze potential anticipated operational occurrences. Transient response during a Loss of Coolant Accident and a Reactivity Initiated Accident were also simulated. This computational study demonstrated that for normal operating conditions, the SiC concept cladding performed as well as the baseline for the standard power cases evaluated. The ramping evaluations indicate potential fracturing of the SiC-SiC composite of the composite cladding compared to the Zircaloy-4 cladding due to the temperature gradient and the subsequent differential thermal conductivity degradation and swelling across the composite thickness. In conclusion, the rod fails early at low enthalpy for RIA but survives a LOCA with minimal material loss due to high temperature steam corrosion.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Methodology for Mechanical Property Testing on Fuel Cladding Using an Expanded Plug Wedge Test

To determine the tensile properties of irradiated fuel cladding in a hot cell, a simple test was developed at ORNL and is described fully in US Patent Application 20060070455, Expanded plug method for developing circumferential mechanical properties of tubular materials. This method is designed for testing fuel rod cladding ductility in a hot cell utilizing an expandable plug to stretch a small ring of irradiated cladding material. The specimen strain is determined using the measured diametrical expansion of the ring. This method removes many complexities associated with specimen preparation and testing. The advantages are the simplicity of measuring the test component assembly in the hot cell and the direct measurement of specimen strain. It was also found that cladding strength could be determined from the test results. The basic approach of this test method is to apply an axial compressive load to a cylindrical plug of polyurethane (or other materials) fitted inside a short ring of the test material to achieve radial expansion of the specimen. The diameter increase of the specimen is used to calculate the circumferential strain accrued during the test. The other two basic measurements are total applied load and amount of plug compression (extension). A simple procedure is used to convert the load circumferential strain data from the ring tests into material pseudo-stress-strain curves. However, several deficiencies exist in this expanded-plug loading ring test, which will impact accuracy of test results and introduce potential shear failure of the specimen due to inherited large axial compressive stress from the expansion plug test. First of all, the highly non-uniform stress and strain distribution resulted in the gage section of the clad. To ensure reliable testing and test repeatability, the potential for highly non-uniform stress distribution or displacement/strain deformation has to be eliminated at the gage section of the specimen. Second, significant compressive stresses were induced by clad bending deformation due to a clad bulging effect (or the barreling effect). The barreling effect caused very large localized shear stress in the clad and left testing material at a high risk of shear failure. The above combined effects will result in highly non-conservative predictions both in strength and ductility of the tested clad, and the associated mechanical properties as well. To overcome/mitigate the mentioned deficiencies associated with the current expansion plug test, systematic studies have been conducted. Through detailed parameter investigation on specific geometry designs, careful filtering of material for the expansion plug, as well as adding newly designed parts to the testing system, a method to reconcile the potential non-conservatism embedded in the expansion plug test system has been discovered. A modified expansion plug testing protocol has been developed based on the method. In order to closely resemble thin-wall theory, a general procedure was also developed to determine the hoop stress $\sigma_\theta$ in the tested ring specimen. A scaling factor called $\chi$-factor is defined to correlate the ring load F ring into hoop stress $\sigma_\theta$, and written as $\sigma_\theta = \chi \frac{F_{ring}} {tl}$. The generated stress-strain curve agrees very well with tensile test data in both the elastic and plastic regions.

expansion plug test↗
DOE OSTI2019

ATF material model development and validation for priority cladding concepts

Development, implementation, and validation of material and behavior models for accident tolerant fuel (ATF) concepts in the Bison fuel performance code began in 2014 in response to the events that occurred at the Fukushima Daichii nuclear power plant in March 2011. Early on the focus was on U 3 Si 2 fuel and FeCrAl cladding as part of a high impact problem through the Nuclear Energy and Advanced Modeling Simulation (NEAMS) program. Then, developments for Cr 2 O 3 -doped UO 2 fuel, and SiC-SiC and Cr-coated zirconium-based claddings began based upon industry interests. In late fiscal year 2018 the Consortium for Advanced Simulation of Light Water Reactors (CASL) took over further ATF work in Bison in support of the Nuclear Regulatory Commission (NRC) engagement. Discussions with the NRC identified their list of priority fuel and cladding concepts, which included Cr 2 O 3 -doped UO 2 and U 3 Si 2 fuels, and Cr-coated zirconium-based and FeCrAl claddings. In particular, the NRC suggested that reports similar in form to NUREG/CR-7024 that was developed for traditional LWR materials UO 2 and zirconium-based claddings (i.e., Zircaloy-4, M5 ® , ZIRLO TM ) be created for the priority ATF concepts. The approach to ATF capability development in Bison since the beginning has been two-fold: (1) empirical correlations and (2) multiscale model development. Both approaches have uncertainty inherent to them. Uncertainty in empirical correlations is bounded by the experimental data upon which with the correlation was developed. Models developed through a multiscale approach have uncertainty associated with the lower length scale calculations and input parameters that must be propagated to the engineering scale model in Bison. In this report, the recommended models, their range of applicability (e.g., temperature, fluence), and associated uncertainty for the NRC priority clad concepts Cr-coated zirconium based clad and FeCrAl are presented in a manner similar to the aforementioned NUREG. In addition, in the absence of experimental data for validation, demonstration cases are presented that investigate the ballooning and burst behavior of Cr-coated and FeCrAl clads as well as tritium permeation in FeCrAl. A portion of the study by Gamble et al. on FeCrAl failure is also revisited with the latest updated models. First, the nominal results are shown using the as implemented models. Then a sensitivity analysis and uncertainty quantification is performed to assess the impact of the uncertainties in the models on the parameters of interest. For the Cr-clad ballooning case the parameter of interest is the time to burst. For the revisited FeCrAl study, the parameter of interest is the pressure at burst.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Post-irradiation Examinations of Accident Tolerant Fuel Candidates with Iron-alloy Cladding

A collaborative effort of the Advanced Fuels Campaign together with industry consortia is focused on the development of enhanced Accident Tolerant Fuels (ATF) for Light Water Reactors (LWRs). The scoping studies, referred to as ATF 1 irradiations, are being performed in the Advanced Test Reactor of Idaho National Laboratory using drop-in style irradiations. The Post Irradiation Examinations (PIEs) of ATF-1 began in 2017 and continued in 2019. This report describes the results of examinations on ATF-1 irradiations designed to test the performance of iron-based cladding. Two rodlets with commercial iron-based cladding and UO2 pellets have been irradiated to low burnup (i.e., < 20 GWd/tHM). The first rodlet had an Alloy33® cladding, while the other had a Kanthal APMT® cladding. The performed PIEs were focused on the determination of fuel microstructure, evaluation of fuel cladding interaction and irradiation-induced variations of cladding mechanical properties. For both rodlets, no significant changes in the cladding hardness were measured, and the cladding hoop strain remained limited. Gamma tomography revealed no Cs axial migration as expected for LWR fuel, but significant cesium radial migration, which was induced by thermal migration consequent to high irradiation temperatures compared to standard LWR irradiation temperatures. Microstructure analyses revealed formation of secondary phases along the pellet rim of samples from both rodlets. No wastage or extensive chemical interaction between the fuel and the internal wall of the cladding was observed, suggesting that the secondary phases accumulated at the pellet rim might be due to formation of Cs-U-O compounds. However, further advanced PIEs are necessary to confirm this hypothesis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Overview of the experiments performed with Russian claddings at MTA EK

The mechanical and corrosion behavior of the Russian-type zirconium cladding alloy E110, predominantly used in VVERs, has been investigated for many decades. The recent commercialization of a new alloy E110G, produced on a sponge zirconium basis, gave the opportunity to compare the mechanical properties of the old E110 and the new E110G fuel claddings. The mechanical experiments performed with samples from both claddings in MTA EK including axial and tangential (ring) tensile tests at 20, 150 and 300 Celsius degrees. The chosen geometries for the tests were modelled using finite element method. Due to the anisotropy of the cladding tubes the axial yield strength was 6% higher than the tangential. The yield strength of the E110G alloy was 12% higher than for the E110 claddings at ambient temperature, but this difference decreased to 5% at 300 C. degrees. Some samples underwent chemical or thermal treatments and the effects of these were evaluated. These treatments were slight oxidation in steam at 800 C. degrees up to 3 ECR%, hydridation at 600 C. degrees up to 400 ppm, and heat treatment in argon atmosphere at temperatures between 600 and 1000 C. degrees. The results confirm that the heat treatment has more significant effect on the yield strength of the claddings than the chemical treatments. While the yield strength of the oxidized samples was slightly higher, when combined with the data from the heat treatment at 800 C. degrees, the oxidation itself seems to have lowered the yield strength as the thickness of the metal decreased. The hydridation of the cladding samples did not change the yield strength and the samples remained ductile even at room temperature. The thermo-mechanical creep of the two cladding alloys was also investigated. Biaxial and uniaxial creep experiments were performed with 100 mm long cladding tubes. In the biaxial test, samples of both alloys were pressurized with argon to 5, 7.5 and 10 MPa at 500 C. degrees and the test lasted till the depressurization of the tubes. The uniaxial creep was measured at 300 and 500 C. degrees in the axial direction and constant stress was applied using weights. This stress was 14.8 and 27.7 MPa at 300 C. degrees and 3.7 and 6.5 MPa at 500 C. degrees. The elongation and diametrical change of the samples were measured using different methods. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Reactivity-initiated accident simulation to inform transient testing of candidate advanced cladding

This paper leverages three-dimensional reactor-core-simulation capabilities to provide information on appropriate experimental test conditions for candidate advanced cladding materials in a control-rod-ejection event. These test conditions use three-dimensional nodal kinetics simulations of a reactivity-initiated accident (RIA) in a representative state-of-the-art pressurized water reactor with both nuclear-grade iron-chromium-aluminum (FeCrAl) and silicon-carbide (SiC-SiC)-based cladding materials. The effort yields boundary conditions for experimental mechanical tests, specifically peak cladding strain during the power pulse following the rod ejection. The impact of candidate cladding materials on the reactor kinetics behavior of RIA progression versus that of reference Zr cladding is predominantly due to differences in -1) fuel mass/volume/specific power density, -2) spectral effects due to parasitic neutron absorption, -3) control rod worth due to hardened (or softened) spectrum, and -4) initial conditions due to power peaking and neutron transport cross sections in the equilibrium cycle cores resulting from hardened (or softened) spectrum. This study shows minimal impact of SiC-SiC-based cladding configurations on the transient response versus reference Zircaloy cladding. However, the FeCrAl cladding response indicates similar energy deposition, but with significantly shorter pulses of higher magnitude. Therefore, the FeCrAl-based cases have a more rapid fuel thermal expansion rate, and the resultant pellet-cladding interaction occurs more rapidly. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

The PROMETRA program: plane strain tests on fresh and highly irradiated zircaloy-4, ZIRLO and M5 fuel claddings

In the framework of the PROMETRA program, clad hardening constitutive laws (stress-strain curves) have been already implemented in the SCANAIR code. These laws were derived from uniaxial hoop tensile tests performed on narrow machined ring samples. However, they do not reproduce the conditions of strain multi-axiality acting on cladding during a RIA (Reactivity-Initiated Accident) in a PWR. As a consequence, a new specimen called Plane Strain Tensile (PST) similar to the annular sample developed by Pennsylvania State University has been developed at CEA on a wider notched ring. It aims to accurately predict the cladding failure in more prototypical conditions, close to plane strain hoop tension in the middle of the gage section. This paper details the PST tests implementation and results in the framework of the PROMETRA program on fresh and irradiated claddings. The development of this test has led also to the development of a technique to measure local deformations in the gage section of the PST specimen using 2-D Digital Image Correlation (DIC). Tests performed on fresh Zircaloy-4 and on irradiated Zircaloy-4, M5 and ZIRLO specimens are analyzed relying of post-test examinations and finite element simulations. In the case of tests performed on irradiated Zry-4 and ZIRLO at 280 and 350 Celsius degrees, the presence of a hydride rim favors the crack nucleation on the outer surface. In this case, rupture nucleation occurs preferentially in the vicinity of notches. Such tests are not representative of the rupture of a cladding subjected to plane strain loading. Nevertheless, they give a good representation of the initiation of rupture observed during RIA on irradiated cladding, resulting in a hydride damage. Zry-4 specimens tested at 480 Celsius degrees and M5TM at 280 and 350 C. degrees result in a ductile fracture nucleation in the center of the specimen. This rupture is prototypical of a ductile rupture of an irradiated specimen subjected to a plane strain loading. However, the nucleation on inner surface by shear at the mandrel contact is not prototypical of a clad rupture during the PCMI (Pellet Clad Mechanical Interaction) phase of a RIA (membrane stresses in the clad thickness in RIA). An important lesson learned from this experimental approach is that the PST test is a particularly complex test, both in its implementation and interpretation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2025

Time-Dependent 600 °C Post-Weld Heat Treatment Response of Laser-Wire-Directed Energy Deposited Austenitic Stainless Steel Claddings on Carbon Steel

Austenitic stainless steel overlay claddings are often used to protect nuclear reactor pressure vessels from corrosion. Arc welding processes are traditionally employed, which require a high heat input to bond the dissimilar metals, creating a large heat-affected zone and high residual stress that demands a lengthy stress-relieving post-weld heat treatment, up to 48 hours. High energy beam (e.g., laser, electron beam) directed energy deposition-based overlay cladding processes show promise for fabricating claddings with superior performance compared to arc welded ones because of their localized heat input; however, their necessity and time of post-weld heat treatment are currently unclear. In this research, we investigate the time dependence of a 600 C stress-relieving post-weld heat treatment of austenitic stainless steel clad on low carbon steel by laser-wire-directed energy deposition, comparing the microstructure and mechanical properties of the as-fabricated condition to 2-, 10-, and 48-hour heat treatment times. Diffusion simulations and microscopy show a gradual carburization of the stainless steel cladding and decarburization of the heat-affected zone/base metal. Nanoindentation testing shows up to a 1.8–2.69 embrittlement in the cladding and a 0.33 to 0.69 softening in the heat-affected zone after the 48-hour post-weld heat treatment. Tensile testing of the cladding composites reveals that increasing post-weld heat treatment time raises clad cracking susceptibility due to intergranular carbide formation and shows that laser-wire-directed energy deposition claddings may not require post-weld heat treatment.

Coating↗
DOE OSTI2022

Critical heat flux on zircaloy and accident tolerant fuel cladding under prototypical conditions of pressurized and boiling water reactors

First of a kind high spatial and temporal resolution temperature measurements of Pressurized Water Reactor and Boiling Water Reactor simulated fuel pins (high heat flux cosine profile heaters) in annular rod type geometry during nucleate boiling, departure from nucleate boiling and re-wetting has been performed. Distributed fiber optic measurements of the cladding axial temperature with spatial resolutions down to 2.5 mm axially at up to three equally spaced azimuthal locations per rod have demonstrated the similarities and differences of the different accident tolerant fuel cladding materials. These experiments show the performance of accident tolerant cladding with a precise measurement of the location and evolution of dryout followed by the re-wetting phenomena (see supplementary time history of a critical heat flux event). Experimental parameters varied coolant mass fluxes from 1695 to 2712 kg/m 2 s, pressures from 10 to 20 MPa, and inlet subcooling from 10 to 55 °C. It was found that the coatings and different material have little effect on the value of the critical heat flux. Any discrepancies between the local critical heat flux verified among the cladding material are attributed to different departure from nucleate boiling phenomena captured during the experiment with the high-resolution optical fiber temperature data. Substantial differences in the performance of the different cladding material to survive a critical heat flux event were however observed by post-test X-ray and optical inspection. External evident damage and internal cracks to the simulated fuel pin having the bare zircaloy cladding were observed in regions associated with the occurrence of the critical heat flux for longer periods. Such damages were associated with zirconium oxidation, and not noticed on the Cr coated zircaloy or FeCrAl claddings. These results indicate that the use of FeCrAl or a Cr coated zircaloy as cladding in Pressurized Water Reactor and Boiling Water Reactor may increase the survivability of the cladding material for short transients.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2023

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

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

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

Structure and tribological properties of steel after non-vacuum electron beam cladding of Ti, Mo and graphite powders

The non-vacuum electron beam cladding technique was used to fabricate layers alloyed with Ti, Mo and C on the surface of low-alloyed steel. Two types of experiments were carried out. In the first experiment, a mixture of Ti and graphite powders was used for cladding; in the second, a mixture of Ti, Mo and graphite powders was used for cladding. CaF{sub 2} powder or a mixture of CaF{sub 2} and LiF powders was used as flux. The thickness of the cladded layers was in the range of 2–2.2 mm. The structure of the layers was studied using optical microscopy, scanning electron microscopy (SEM), transmission electron microscopy (TEM), and X-ray diffraction (XRD). The microhardness after cladding of the layers fabricated by cladding of Ti and graphite powders was 8–9 GPa, while the microhardness of layers with Mo additions reached 11–12 GPa. The highest wear resistance at sliding friction and friction in abrasive environment was reached in the samples fabricated using Ti, Mo and graphite mixture due to the higher hardness and the martensite–austenite structure of the matrix. The wear resistance against fixed abrasive particles was 2.4 times higher compared to that of carburized and quenched steel. - Highlights: • Ti, C and Mo mixture of powders was cladded using non-vacuum electron beam treatment. • The depth of the cladded layers was 2.0 … 2.2 mm. • The microhardness of layer with Mo, Ti and C additions reached ~ 11 … 12 GPa. • The hardening of the layers caused by the formation of TiC particles and martensitic matrix • Wear resistance of cladded coatings was 2.4 higher than carburized steel.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Development of a thermal creep model for aluminum alloy 6061 cladding in U-10Mo monolithic fuel plates

Plate-type fuel elements consisting of a high-density, low-enriched uranium (LEU) U–10Mo-based fuel foil encapsulated in an aluminum alloy (AA) cladding are fabricated using the hot isostatic pressing (HIP) technique. During the HIP process, the fuel plate system is heated to 560 °C, then cooled to room temperature. This heat cycle significantly affects the mechanical properties of the aluminum cladding, and experimental investigations have shown that, post-HIP bonding, the mechanical properties of the aluminum cladding transition from those of AA 6061-T6 to something closer to the O temper. More specifically, the ultimate strength of the cladding decreases while its ductility increases, making it challenging to capture the changes in mechanical behavior and material properties. Understanding the residual stresses generated during the HIP process is critical for assessing the fuel plate’s integrity under various temperature, pressure, and irradiation. To simulate the HIP bonding process, the elastic, plastic, and thermal properties of the cladding are assumed to be similar to those of AA 6061-O temper. However, the primary challenge lies in the lack of available data for the creep model of the AA 6061 cladding during this transient process of HIP. The present study focuses on developing a computational model that predicts the creep behavior of the aluminum cladding in the fuel plates during the HIP process, as cladding creep significantly influences the residual stresses generated in U-10Mo fuel plates during HIP fabrication. Furthermore, as HIP bonding occurs at high temperatures that are nearing the melting point of aluminum, the present work considered a temperature-dependent Arrhenius-type creep model. In particular, a hyperbolic sine creep model is employed to estimate the creep properties of the as-fabricated aluminum cladding. In conclusion, the residual stresses predicted in the U-10Mo fuel when using the newly calibrated creep model closely align with the experimental measurements, validating the model’s accuracy.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2013

Development of Diffusion Barrier Coatings and Deposition Technologies for Mitigating Fuel Cladding Chemical Interactions (FCCI)

The goal of this project is to develop diffusion barrier coatings on the inner cladding surface to mitigate fuel-cladding chemical interaction (FCCI). FCCI occurs due to thermal and radiation enhanced inter-diffusion between the cladding and fuel materials, and can have the detrimental effects of reducing the effective cladding wall thickness and lowering the melting points of the fuel and cladding. The research is aimed at the Advanced Burner Reactor (ABR), a sodium-cooled fast reactor, in which higher burn-ups will exacerbate the FCCI problem. This project will study both diffusion barrier coating materials and deposition technologies. Researchers will investigate pure vanadium, zirconium, and titanium metals, along with their respective oxides, on substrates of HT-9, T91, and oxide dispersion-strengthened (ODS) steels; these materials are leading candidates for ABR fuel cladding. To test the efficacy of the coating materials, the research team will perform high-temperature diffusion couple studies using both a prototypic metallic uranium fuel and a surrogate the rare-earth element lanthanum. Ion irradiation experiments will test the stability of the coating and the coating-cladding interface. A critical technological challenge is the ability to deposit uniform coatings on the inner surface of cladding. The team will develop a promising non-line-of-sight approach that uses nanofluids . Recent research has shown the feasibility of this simple yet novel approach to deposit coatings on test flats and inside small sections of claddings. Two approaches will be investigated: 1) modified electrophoretic deposition (MEPD) and 2) boiling nanofluids. The coatings will be evaluated in the as-deposited condition and after sintering.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Evaluation of Tritium Content and Release from Surry-2 Fuel Cladding (Milestone Report)

To design and operate future reprocessing plants in a safe and environmentally compliant manner, the amount and form of tritium in the used nuclear fuel (UNF) must be understood and quantified. To gain a better understanding of how tritium in cladding will behave during processing, scoping tests are being performed to determine the tritium content in the cladding pre- and post-tritium pretreatment. A sample of Surry-2 pressurized water reactor (PWR) cladding was heated to 1100–1200°C to oxidize the zirconium and release all of the tritium in the cladding sample. The tritium content was measured to be ~240 µCi/g. Cladding samples were heated to 500ºC, which is within the temperature range (480 - 600ºC) expected for standard air tritium pretreatment systems, and to a slightly higher temperature (700ºC) to determine the impact of tritium pretreatment on tritium release from the cladding. Heating at 500°C for 24 hr removes ~0.2% of the tritium from the cladding, and heating at 700°C for 24 hr removes ~9%. Thus, a significant fraction of the tritium remains bound in the cladding and must be considered in operations involving cladding recycle.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Fatigue Testing and Characterization of Pre-hydrided Zircaloy-4 Cladding Tubes

The solubility of hydrogen in Zircaloy-4 (Zry-4) at the cladding operating temperature is near 100 wt. ppm. Above this solubility limit, excess hydrogen precipitates as δ-hydride platelets in the cladding material. Because of the effect of the thermal gradient across the cladding thickness on the migration and precipitation of hydrogen, hydride rims are often observed at the cladding outer surface; under excessive corrosion conditions hydride blisters are possible. It is well known that the fatigue of metal alloys, especially high cycle fatigue, is sensitive to the status of surface including the microstructure, roughness, and residual stress. Thus, the question considered herein, is whether excessive hydrogen pickup modifies the microstructure such that it has a degradation effect on the fatigue performance of cladding during operation. This report describes the evaluation of fatigue performance of a pre-hydrided Zry-4 cladding. A commercial Zry-4 was polished and pre-hydrided to 800 ppm and 1300 ppm H contents. The fatigue testing was conducted under strain control at 5 Hz with fully-reversed bending by using a cyclic integrated reversible bending fatigue tester (CIRFT). Six specimens with 1300 ppm and one specimen with 800 ppm were tested. Significant variation in test results were observed. While two of the 1300 ppm specimens failed with less than 1 ×10 5 cycles (at 0.32% and 0.41%), the four other samples did not fail over the strain amplitude range of 0.25% to 0.38%. Interestingly, the fracture initiation site of the failed samples was on the outside diameter (OD) surface rather than on the inside diameter (ID) surface as is typical for the as-polished cladding. This suggested a degrading effect of the hydriding process. Subsequently, three of the unfailed specimens were then further tested at ~0.43% to observe where failure initiation occurred. Significant variation was also observed in these three specimens. One specimen failed at ~9000 cycles while the two others failed at 43000 and 1.06 ×10 5 cycles. The performance here correlated to the location of failure initiation; failure initiated on the OD in the sample that failed after 9000 cycles while it initiated on the ID in the sample that failed after 43000 and 1.06 ×10 5 cycles. Flat features at the OD initiation sites suggest a brittle hydride feature on the surface of those samples was the cause of the degradation in fatigue performance, though the overall hydrogen levels in all samples was similar. A summary of all the observations is provided below: • The un-failed specimens with 1300 ppm H were cycled to failure with a higher amplitude near 0.43%. In addition, the fatigue-treated specimen tended to have a longer fatigue at the same induced amplitude. • Fractography revealed a mixed failure mode for the pre-hydrided specimen. Particularly, the specimen with fracture initiation site (FIS) located on the outer diameter surface of tube tended to have a shorter fatigue life than that of FIS on the inner diameter surface. • Etched cross section was shown to have hydride platelets aligned with tube longitudinal axis as expected, and the density of hydrides is at the similar level as in literature data. Meanwhile, a LECO procedure was applied for hydrogen concentration measurement, which showed the measured hydrogen contents are close to the nominal value. • With the polished cladding tube as baseline, O’Donnell-Lager (O-L) analysis showed that a decrease of about 50% in reduction-of-area (RA) would be needed for the O-L fitting to the fatigue data of 1300 ppm cladding. The suggestion for the next steps is provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Transmission of laser pulses with high output beam quality using step-index fibers having large cladding

An apparatus and method for transmission of laser pulses with high output beam quality using large core step-index silica optical fibers having thick cladding, are described. The thick cladding suppresses diffusion of modal power to higher order modes at the core-cladding interface, thereby enabling higher beam quality, M.sup.2, than are observed for large core, thin cladding optical fibers. For a given NA and core size, the thicker the cladding, the better the output beam quality. Mode coupling coefficients, D, has been found to scale approximately as the inverse square of the cladding dimension and the inverse square root of the wavelength. Output from a 2 m long silica optical fiber having a 100 .mu.m core and a 660 .mu.m cladding was found to be close to single mode, with an M.sup.2=1.6. Another thick cladding fiber (400 .mu.m core and 720 .mu.m clad) was used to transmit 1064 nm pulses of nanosecond duration with high beam quality to form gas sparks at the focused output (focused intensity of >100 GW/cm.sup.2), wherein the energy in the core was <6 mJ, and the duration of the laser pulses was about 6 ns. Extending the pulse duration provided the ability to increase the delivered pulse energy (>20 mJ delivered for 50 ns pulses) without damaging the silica fiber.

42 ENGINEERING↗