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At least 109 records · Page 6
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Nd, SbNd and Sb 3 Nd 4 and their interactions with the cladding alloy HT9

Lanthanide fission products, such as neodymium, formed during the irradiation of metallic fuels are known to cause deleterious effects from chemical interactions occurring at the fuel-cladding interface; a phenomenon known as fuel-cladding chemical interaction (FCCI). The use of fuel-based additives that bind with the lanthanide elements within the fuel meat, alleviating their interactions at the fuel-cladding interface, is one potential method proposed to mitigate the FCCI phenomenon and extend the burnup potential of such metallic fuel systems. In this study, antimony (Sb) is evaluated as one such additive, and neodymium (Nd) is used to represent the lanthanides. A Sb-Nd alloy is fabricated, which consists of two intermetallic phases, SbNd and Sb 3 Nd 4 . Isothermal diffusion couple experiments are carried out at 675 °C for 24 h between Nd and the Fe-12Cr based HT9 cladding alloy. The results are compared against similar diffusion couple experiments carried out between the Sb-Nd alloy and also HT9. Inter-diffusion between Nd/HT9 diffusion couple is characterized and the phases Fe 17 Nd 5 and Fe 17 Nd 2 were found to form whereas no interactions were observed in the Sb-Nd/HT9 diffusion couple. The lack of compound forming tendencies between Fe, the primary alloying constituent of HT9, and SbNd is elucidated through density functional theory (DFT) calculations on enthalpy of mixing, aligning well with the experimental observations. Lastly, the strong binding of Nd with Sb appears to be favorable, alleviating Nd interactions with HT9 constituent elements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Cladding and Butt-Joining Dissimilar Aluminum Alloys Simultaneously via Shear Assisted Processing and Extrusion

Shear assisted processing and extrusion has been used to fabricate multi-metallic tubing that incorporates regions with clad and butt-joined structures. In this work, aluminum alloys 6061, 7075, and 1100 are combined to create a monolithic extruded tubing with weldable ends (6061), a high strength center (7075), and a corrosion-resistant cladding (1100). Optical microscopy shows a gradual, defect-free, transition between alloys at the butt-joint interfaces and a sharp transition between alloys at the clad interface. Tensile testing proved the mechanical coherency of the joint, with failure occurring outside the tri-metallic structure. In conclusion, smoothed particle hydrodynamic simulations identified a spiral pattern of the cladding alloy that leads to inconsistent cladding thickness, and identified a simplified 2-alloy combination of aluminum alloys 6082 and 2014 that accomplished the design objectives of the tri-metallic structure.

Coatings↗
DOE OSTI2021

Transient testing of uranium silicide fuel in zircaloy and silicon carbide cladding

Using the newly restarted Transient Reactor Test (TREAT) facility at Idaho National Laboratories, fresh (unirradiated) U 3 Si 2 fuel pellets were tested under reactivity-initiated accident conditions in dry helium capsules. These tests were performed with total energy depositions sufficient to induce melting of the U 3 Si 2 fuel pellets based on available thermophysical property data. The instrumented experiments utilized in-pile thermocouples and infrared pyrometers to measure the cladding temperature during the transient, which are in reasonable agreement with predicted peak cladding temperatures using a simple BISON fuel performance model. This same model is then used to predict expected cladding stresses during the transients in areas where pellet cladding interaction is expected. It was shown that for transient energy depositions up to 528J/g rod like geometry of the test articles was maintained. While it is acknowledged that more confirmatory testing is required in more prototypic environments, these first of their kind experiments have demonstrated that unirradiated U 3 Si 2 fuel and unirradiated SiC-SiC claddings can likely retain a “rod like” geometry for transient energy depositions up to 528J/g.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Semi-integral LOCA test of cold-spray chromium coated zircaloy-4 accident tolerant fuel cladding

Enhancing the accident tolerance of PWRs came under serious consideration following the 2011 Fukushima Daiichi nuclear power plant events. While the current nuclear fuels are based on mature technology and have an excellent operational record, further enhancing their strength through accident tolerant fuel (ATF) research and development continues to be of interest. Particularly, the chromium coated clad ATF technology represents an evolutionary change to improve the high temperature performance of existing Zircaloy claddings. Additionally, despite several years of investigation of coated cladding, the integral loss of coolant accident (LOCA) performance of chromium coated cladding has had little attention. In this work, cold-spray chromium coated zircaloy-4 claddings are evaluated using EDF's semi-integral LOCA test facility. Results show that the chromium coating induces a reduced ballooning and burst and an increased burst temperature in the investigated conditions. However, these tests exhibit limited benefits concerning the post-quench behavior of the coated specimens which occurs in the last stage of a prototypical LOCA scenario.

36 MATERIALS SCIENCE↗
DOE OSTI2022

A study on texture stability and the biaxial creep behavior of as-hydrided CWSR Zircaloy-4 cladding at the effective stresses from 55 MPa to 65 MPa and temperatures from 300°C to 400°C

The creep rupture of high burnup used nuclear fuel (> 45 GWD/MTU) cladding is regarded as one of the failure mechanisms during long-term dry storage. A high amount of zirconium hydride in the cladding matrix would degrade the mechanical properties of the cladding especially leading to delayed hydride cracking. To better understand the influence of zirconium hydride on the biaxial thermal creep performance and the crystalline texture stability of Zircaloy-4 cladding, the pressurized tube technique is applied to test the durability of the as-hydrided material. Tests were performed on as-received Zircaloy-4 tubular specimens as well as-hydrided ones with targeted 300 wt parts per million (wppm) or 750 wppm hydrogen. The biaxial creep experiments were conducted at temperatures from 300 degrees C to 400 degrees C and at equivalent stresses at mid-wall from 55 MPa to 65 MPa. The hydridation process prior to creep tests induces the formation of FCC delta-hydride platelets along the circumferential direction of the tube. This alignment and phase structure of hydride show no significant change after biaxial creep tests. The creep strain-rate negatively depends on the hydrogen content. The synchrotron wide-angle X-ray diffraction (WAXD) technique and electron backscatter diffraction (EBSD) analysis were applied for the study of the crystallographic orientation relationship. zirconium-to-hydride grain orientation follows Shoji-Nishiyama crystallographic relationship. This relationship is stable before and after creep deformation. These results of creep performance and texture stability of Zircaloy-4 claddings can help support the design basis of interim and long-term dry storage facilities.

creep↗
DOE OSTI2022

BISON fuel performance modeling optimization for experiment X447 and X447A using axial swelling and cladding strain measurements

With the recent need to qualify new reactor designs such as the Versatile Test Reactor (VTR), fuel performance calculations need to be performed to determine safety criteria of the proposed designs. In order to validate the fuel performance results obtained by a fuel performance code, BISON, for new reactor designs, legacy fuel from EBR-II and FFTF MFF with Post -Irradiation Examination (PIE) data need to be used as validation cases to benchmark models. Here in this work, BISON has been paired with the Fuels Irradiation & Physics Database (FIPD) and IFR Materials Information System (IMIS) to supply PIE data for comparison with simulations of EBR-II experiments X447/X447A. X447/X447A were assessed by implementing models for Fuel Cladding Chemical Interaction (FCCI) within BISON and optimizing the friction coefficient between the fuel surface and the cladding, the anisotropic swelling factor, and the HT9 first thermal creep scalar (which scales the first term in the HT9 creep equation) to best match the PIE axial fuel swelling height and cladding profilometry for all pins in X447/X447A. The optimal values were found using a generic algorithm developed to select different values for the three parameters until end criteria was met and error couldn’t be reduced further. The BISON-simulated cladding profilometry was evaluated using Standard Error of the Estimate (SEE) to account for the profile shape of the cladding profilometry. Optimal values for the friction coefficient, anisotropic fuel swelling factor, and HT9 first thermal creep scalar were found to best fit the BISON simulation results to the PIE measurements found in IMIS and FIPD. Improvements to current models are suggested to account for the underprediction of fuel swelling at low burnups and the overprediction of fuel swelling at higher burnups observed for the axial fuel swelling height. Although two pins in EBR-II X447/X447A (DP70 and DP75) were known to fail due to FCCI, none of the pins simulated in BISON reached a cumulative damage fraction (CDF) above 0.008 with FCCI correlations coupled in the BISON simulations. The error estimate generated for all pins in X447/X447A using optimal values was 209 µm, which is deemed acceptable.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Model for determining rupture area in Zircaloy cladding under LOCA conditions

The dispersal of nuclear fuel through a cladding rupture opening during a transient event is the primary safety concern preventing the United States nuclear industry from extending burnup beyond a peak rod average burnup of 62 GWd/tU. The majority of research thus far has prioritized understanding, or at a minimum prioritized quantifying, fuel fragmentation and pulverization as a function of burnup and to a lesser extent onset temperature. Early high burnup loss-of-coolant accident tests performed at the Halden Boiling Water Reactor evaluated the impact of relocation on cladding temperature, whereas historical and recent cladding burst tests focused primarily on generating data related to cladding balloon and burst timing. However, very little effort has been put forth to develop a relationship for determining the rupture opening dimensions for Zircaloy cladding tubes bursting during a simulated light water reactor loss-of-coolant accident. This is significant because high burnup loss-of-coolant accident data suggest dispersal could be limited by the size of the rupture opening or rupture dimensions, thereby minimizing the safety concern. Therefore, the purpose of this research is to build a database composed of all simulated loss-of-coolant accident data to develop an empirical model that conservatively captures the rupture opening dimensions and translates those dimensions to rupture area. Analysis of the experimental data indicates the rupture dimensions are strongly correlated to the peak balloon strain. This relationship is leveraged to develop a rupture length empirical correlation as a function of peak balloon strain and a rupture width empirical correlation as a function of rupture length. Outliers in the experimental data are identified and discussed. Finally, the rupture opening empirical model is presented in such a way for general implementation into high fidelity fuel performance codes or fuel performance codes used for safety analyses.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Mechanistic modeling of lifetime distribution of SiC/SiC composite claddings

Silicon carbide (SiC) fiber-reinforced SiC matrix (SiC/SiC) composites have emerged as a new material candidate for fuel claddings in light water reactors. Recent studies showed that the load capacity of SiC/SiC materials exhibits a considerable statistical variation. Therefore, reliability analysis plays a critical role in design of SiC/SiC composite claddings. Here, this paper presents a probabilistic model for the lifetime distribution of SiC/SiC composites. The model is anchored by a multiaxial stress-based failure criterion and subcritical damage accumulation mechanism. Based on the kinetics of subcritical damage growth, the lifetime distribution of a laboratory test specimen for any given loading history can be calculated. A finite weakest-link model is used to extrapolate the lifetime distribution of test specimens to full-length claddings. It is shown that the damage accumulation mechanism has a strong influence on the lifetime distribution of the cladding. This finding highlights the importance of understanding the static fatigue behavior of SiC/SiC composites. The present analysis also demonstrates an intricate length effect on the failure probability of the cladding, which is expected to play a crucial role in design extrapolation.

36 MATERIALS SCIENCE↗
DOE OSTI2022

In-situ thermomechanical measurements during burst testing of Zr claddings using digital image correlation and infrared thermography techniques

Burst testing has long been a relied-on method for assessing fuel cladding materials in light-water reactoraccident scenarios. However, burst testing historically provided minimal in-situ observations beyond a few thermocouple readings, pressure from transducers, and post-test deformation measurements. Acquisition of full cladding thermomechanical data throughout a burst test would be invaluable to fuel system qualification, material model development and fuel performance code validation, such as the BISON fuel performance code. Recently, methodologies for simultaneously applying 2-dimensional digital image correlation and infrared thermography techniques to quantify the relationship between temperature, internal cladding pressure, and deformation during burst testing were developed. This work details results from the application of these techniques to burst testing of Zircaloy-4 cladding segments subjected to loss-of-coolant accident transients in an air environment. Two types of tests were performed with: (a) cladding segments connected to a fixed pressure reservoir as well as (b) closed systems where internal pressure increased during transients. In-situ thermomechanical measurements including strain and thermal gradient data are presented, demonstrating the effectiveness of applying combined surface measurement techniques to burst testing

Bell, Sam↗
DOE OSTI2022

A review of cladding failure thresholds in RIA conditions based on transient reactor test data and the need for continued testing

Transient reactor experiments on light water reactor (LWR) fuel pins had been conducted since the beginning of the nuclear era to help determine core coolability and cladding failure thresholds. During one such test in November of 1993 at the CABRI transient test reactor on a test involving a high burnup fuel rod with a corroded Zircaloy-4 cladding it was first observed that cladding failures could occur prior to a departure from nucleate boiling (pre-DNB) at lower-than-expected peak radial average enthalpies. This paper will present an independent review of the publicly available transient reactor test database on higher burnup LWR pins conducted at the CABRI and NSRR reactors as well as review of a selection of published out of pile mechanical testing methods. The purpose of the review is to determine how well the new regulatory limits are supported by experimental data. The review will identify if additional transient reactor tests could provide additional support for the NRC guidance or identify the need for revisions. The evaluation will consider how far the existing database can be extrapolated when considering low hydrogen zirconium alloy claddings (with and without protective coatings) containing very high burnup (> 70 MWd/kgU) UO2 fuel pellets. Finally, the authors will suggest how out of pile mechanical tests can be used in conjunction with a limited number of transient reactor tests to develop cladding specific failure thresholds in RIA type transients.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Nano-porous GaN cladding and scattering loss in edge emitting laser diodes

We report continuous wave operation of electrically injected InGaN laser diodes using nano-porous GaN n-side cladding with 33% porosity. At 454 nm emission wavelength, the pulsed injection slope efficiency is 0.24 W/A with a high loss of 82 cm -1 . The considerable 60 cm -1 of excess loss of the nano-porous clad lasers is attributed to scattering at pores in unintentionally 3% porosified layers, supported by numerical modeling. Simulations comparing porous GaN cladding to AlInN cladding for lasers operating at 589 nm indicate that the porous cladding provides similar internal loss and lower thermal impedance.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Plane-Strain Tension Testing of ATF FeCrAl Cladding (Progress Report)

Performance of advanced accident-tolerant fuel (ATF) claddings must be assessed under all relevant scenarios, including normal and off-normal operating conditions, as well as during a reactivity-initiated accident (RIA). During RIA, the fuel pellet may expand and cause significant loads on cladding, a condition known as pellet cladding mechanical interaction (PCMI), which may even lead to cladding failure. This report describes the fiscal year (FY) 2020 research activities and results from the development of separate-effects testing for ATF cladding candidate iron-chromium-aluminum (FeCrAl) tube specimens under loading conditions expected during RIA/PCMI events. To assess material performance under PCMI, a high strain rate burst test, often termed the modified burst test (MBT), is commonly employed. However, execution of such a test in a hot cell requires significant research. To overcome the complexities of performing MBTs in a hot-cell environment on irradiated samples, a plane-strain tension (PST) test was proposed by several authors as an alternative approach for PCMI testing. A new fixture was developed so that PST specimens could be tested using a more accommodating design for hot-cell manipulators. This report presents the PST fixture design, evaluates the possibility of modifying the specimen geometry, and discusses experimental results with unirradiated Al 6061 and FeCrAl tubes at various lengths. Furthermore, this report discusses the relevance and compatibility of the PST results to RIA scenarios and compares the PST and MBT results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Impact of Coating Defects on Performance of Coated Zirconium Cladding

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

36 MATERIALS SCIENCE↗
DOE OSTI2023

Impact of LWR Assembly Structural Features on Cladding Burst Behavior under LOCA Conditions

This report builds on the previous work by assessing the impact of assembly structural features on cladding balloon and burst behavior. The BISON fuel performance code was used to generate two dimensional RZ meshes containing structural features commonly used in nuclear fuel assemblies. The 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 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. Consideration was also given to the effect of rod initial pressure. Overall, the BISON simulations indicated the spacer grids can locally influence cladding performance, however, the impact on the cladding performance was minor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Review and Summary of Corrosion Behavior for Aluminum-Clad Spent Nuclear Fuel in Dry Storage

A description of the corrosion behavior of aluminum alloys used for cladding on aluminum-based, aluminum-clad nuclear fuel used in research reactors under potential dry storage conditions has been compiled. An evaluation is made of potential additional corrosion with water postulated to not be removed during drying. The relative humidity (RH) produced by typical dryness criteria for industrial spent nuclear fuel drying is expected to be low, particularly at high temperatures (17% RH at 20°C and lower at higher temperatures). Existing corrosion data suggests that negligible vapor-phase corrosion of aluminum is expected at such low humidity, even for nominally bare aluminum surfaces. That is, the predicted relative humidity from free water (e.g., 17% RH) is below the “critical” relative humidity (~40% RH at room temperature), below which no significant corrosion is observed for bare aluminum. Furthermore, reported room-temperature corrosion rates are very low even under saturated (100% RH) water vapor. Existing results showing significant vapor-phase corrosion corresponded specifically to conditions of high temper ed with high relative humidity. A relatively large reservoir of (chemically bound) water exists in the aluminum (oxy)hydroxide films on the SNF cladding surface. The estimated total could saturate the gas even at relatively high temperature if fully released; however, its release as molecular water is expected to only be plausible if the temperature during storage exceeds both the drying temperature and the threshold for thermal decomposition of the trihydroxides (~220°C). Therefore, the combination of high temperature and high relative humidity that could drive significant corrosion is considered implausible during sealed dry storage following an appropriate drying process, to include >220°C drying for canisters that may approach or exceed this temperature during storage. Vapor-phase testing of aluminum samples with an adherent (oxy)hydroxide layer, prepared in liquid water to resemble those on actual aluminum-clad spent nuclear fuel (ASNF), did not observe evidence of additional corrosion even at combined high-temperature (up to 180°C) and high-humidity (up to 100% RH) conditions. Instead, small net mass losses were observed for most specimens and attributed to dehydration of the samples, which had been air-dried only prior to testing. ASNF being moved to dry storage is expected to have an existing (oxy)hydroxide film, suggesting that the cladding will be less prone to additional corrosion than bare aluminum metal. If significant corrosion did occur during sealed storage, the overall extent and impact is expected to be small. Corrosion post-closure of a canister would not alter the total amount of hydrogen in the canister, so it would have no effect on the maximum H 2 release already assessed in existing bounding calculations. In addition, the total amount of water in a 25-µm dense bayerite film would consume a only ~8 µm additional aluminum metal on average, if fully consumed by oxidizing aluminum metal to Al 2 O 3 . These conclusions are consistent with ASNF-in-canister simulations to date, which included a reaction pathway for corrosion using kinetics from previous literature and believed to be conservative and predicted very low rates of corrosion.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Effects of Internal and External Heat Sources on Cladding Microstructure and Rupture Performance

In the event of a Loss of Coolant Accident (LOCA), the primary supply of cooling water for a nuclear reactor is lost, leading to a significant pressure differential across the cladding wall. Without adequate cooling, the fuel rods continue to heat as a result of fission reactions. Research at Oak Ridge National Laboratory’s (ORNL’s) Severe Accident Test Station (SATS) is currently focused on evaluating fuel cladding performance using an external infrared lamp as a heat source, whereas legacy testing primarily utilized an internal heating approach. While external heating may better simulate the effect of neighboring fuel rods heating a central rod, it may not accurately represent the internal heat absorption from the fuel during accident transients. The heating dynamics depend greatly on the fuel rod's position within the bundle and the reactor. To fully understand the implications of a LOCA event and assess the influence of internal heating on cladding performance, combined internal heating and pressurization capability was developed at ORNL. Tests were conducted to compare cladding segments heated internally (representing heat from the fuel within the rod) and externally (representing heat from adjacent fuel rods). The findings indicate that both internal and external heating result in comparable rupture temperatures during 5°C/s laboratory LOCA tests and also agree with legacy test data. Axial temperature gradients and internal heat source dispersal were found to significantly impact cladding deformation and rupture geometry. Clear modifications were outlined to further improve the capability with heating rates above 5°C/s.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Effects of Internal and External Heat Sources on Cladding Microstructure and Rupture Performance

In the event of a Loss of Coolant Accident (LOCA), the primary supply of cooling water for a nuclear reactor is lost, leading to a significant pressure differential across the cladding wall. Without adequate cooling, the fuel rods continue to heat as a result of fission reactions. Research at Oak Ridge National Laboratory’s (ORNL’s) Severe Accident Test Station (SATS) is currently focused on evaluating fuel cladding performance using an external infrared lamp as a heat source, whereas legacy testing primarily utilized an internal heating approach. While external heating may better simulate the effect of neighboring fuel rods heating a central rod, it may not accurately represent the internal heat absorption from the fuel during accident transients. The heating dynamics depend greatly on the fuel rod's position within the bundle and the reactor. To fully understand the implications of a LOCA event and assess the influence of internal heating on cladding performance, combined internal heating and pressurization capability was developed at ORNL. Tests were conducted to compare cladding segments heated internally (representing heat from the fuel within the rod) and externally (representing heat from adjacent fuel rods). The findings indicate that both internal and external heating result in comparable rupture temperatures during 5°C/s laboratory LOCA tests and also agree with legacy test data. Axial temperature gradients and internal heat source dispersal were found to significantly impact cladding deformation and rupture geometry. Clear modifications were outlined to further improve the capability with heating rates above 5°C/s.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Magnetic Field Sensing via Acoustic Sensing Fiber with Metglas® 2605SC Cladding Wires

Magnetic field sensing has the potential to become necessary as a critical tool for long-term subsurface geophysical monitoring. The success of distributed fiber optic sensing for geophysical characterization provides a template for the development of next generation downhole magnetic sensors. In this study, Sentek Instrument’s picoDAS is coupled with a multi-material single mode optical fiber with Metglas® 2605SC cladding wire inclusions for magnetic field detection. The response of acoustic sensing fibers with one and two Metglas® 2605SC cladding wires was evaluated upon exposure to lateral AC magnetic fields. An improved response was demonstrated for a sensing fiber with in-cladding wire following thermal magnetic annealing (~400 °C) under a constant static transverse magnetic field (~200 μT). A minimal detectable magnetic field of ~500 nT was confirmed for a sensing fiber with two 10 μm cladding wires. The successful demonstration of a magnetic field sensing fiber with Metglas® cladding wires fabricated via traditional draw processes sets the stage for distributed measurements and joint inversion as a compliment to distributed fiber optic acoustic sensors.

Dejneka, Zach (ORCID:0000000179415708)↗