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28 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Enhanced Stress Relaxation Behavior via Basal ⟨a⟩ Dislocation Activity in Zircaloy-4 Cladding

Herein this work evaluates the stress relaxation behavior of textured Zircaloy-4 cladding to understand how mechanical anisotropy influences pellet-cladding interactions. Uniaxial and biaxial stress relaxation tests are performed using full-tube axial tension and internal pressurization, respectively, aiming to achieve 0.25%, 1%, and 2% equivalent strains in the cladding samples at a temperature of 300°C. Internal pressure relaxation test results display enhanced stress relaxation compared to axial testing results, particularly for samples loaded beyond yield. Results of electron backscatter diffraction indicate increased deformation microstructure during loading and increased strain homogenization and recovery during relaxation for samples loaded via internal pressurization. Analysis indicates that the increased production and activity of basal ⟨a⟩ dislocations play a significant role in the enhanced relaxation measured in samples subjected to internal pressurization.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

$\mathrm{UPDATED}$ $\mathrm{U3SI2}$ thermal creep model and sensitivity analysis of the $\mathrm{U3SI2-SIC}$ accident tolerant $\mathrm{FUEL}$

U 3 Si 2 is a candidate accident tolerant fuel (ATF) replacement for UO 2 . U 3 Si 2 ’s high uranium density and high thermal conductivity are favorable properties in steady-state and accident conditions. Low power performance of this U3Si2-SiC concept fuel is compared to that of UO 2 -Zr4 fuels by implementing models that describe the properties of U 3 Si 2 and SiC-SiC into Idaho National Laboratory's (INL) fuel performance code, BISON. Included in these material models is a thermal creep model for U 3 Si 2 based on compressive creep data. The simulated results are in keeping with community knowledge that the U 3 Si 2 -SiC concept fuel may serve as a replacement for UO 2 -Zr4 fuels during steady-state operation, provided the mSiC layer remains under compression. Through a moderate power history and three 24-month fuel cycles, the mSiC layer remains under compressive stress through a burnup of 80 MWd/kgU. During low power operation, failure of the mSiC layer generally occurs prior to significant thermal creep in U 3 Si 2 . Generally, U 3 Si 2 creep is temperature sensitive and of little importance at the temperatures and stresses simulated during steady operation and during fuel-to-cladding contact. A parameter variation study including 11,520 individual simulations with variations in nominal fuel thermal creep rate, cladding thermal conductivity, cladding irradiation creep and swelling, cladding gap size, and cladding thickness demonstrated that research priorities for this ATF should revolve around reducing cladding thickness as a means to minimize cladding failure. Generally, despite advances in SiC-SiC compliance, the brittle nature of mSiC excludes U 3 Si 2 -SiC for use where fuel cladding contact may occur.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Multidimensional modeling of fuel-cladding friction in an LWR fuel rod

Solving finite element problems with friction often increases the level of difficulty to obtain properly converged solutions. This type of challenge becomes more salient when advanced, possibly multiscale, material models are employed to capture the thermomechanical behavior of fuel and cladding materials. Here the present work details our recent developments in a nuclear fuel performance finite element code for the systematic consideration of friction in nuclear reactor finite element simulations. We show the application of friction and its effects on the mechanics of light-water reactor rods accounting for various fuel constitutive modeling techniques, model dimensionalities, pellet assumed geometries, and power conditions. In particular, we focus on the fuel rod mechanical behavior as it relates to fuel constitutive models, sensitivity to the coefficient of friction, pellet states of stress, and rod elongation. We discuss the trade-offs between the various multidimensional modeling options and highlight the relevance of frictional effects in the prediction of the fuel rod deformation and interfacial stresses. To relate our modeling results with actual reactor operation, simulations including frictional effects are compared with fuel rod elongation experimental data and some challenges for carrying out a full validation of the axial mechanics are discussed.

42 ENGINEERING↗
DOE OSTI2024

Anisotropic Effects of Loading on Zircaloy Cladding Performance

This presentation provides recent results related to the mechanical behavior of Zircaloy cladding for the Light Water Reactor Fuels Workshop/CRAFT Initiative Update presented by the Advanced Fuels Campaign. This presentation focuses on Zircaloy cladding response to pellet-cladding interactions and how mechanical anisotropy makes the cladding sensitive to the testing mode. Results are compared to existing material models used by BISON, and implications for reactor performance are briefly discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Mechanical Response of Cr-Coated Zircaloy-4 Cladding to Pellet-Cladding Interactions

Chromium-coated zircaloy is a promising accident tolerant fuel (ATF) cladding design which can help mitigate high temperature oxidation. The mechanical response of the cladding during and after loading may be altered by the Cr coating throughout anticipated operational and accident occurrences such as a pellet-cladding interaction (PCI). In this work, the mechanical behavior of Cr-coated and uncoated cold worked stress-relieved (CWSR) Zircaloy-4 (Zry-4) cladding to PCI loading conditions are studied. Stress relaxation tests are performed on CWSR Zry-4 cladding using full-tube axial tension and internal pressurization to investigate mechanical anisotropy and rate-limiting deformation mechanisms of the uncoated Zry-4. These results are compared to internal pressure testing results of similar cladding with physical vapor deposition (PVD) Cr-coating to reveal how the Cr-coating affects the cladding mechanical response. Findings are discussed in the context of ATF cladding performance during operation and storage.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Status Report on Design of In-situ Thermomechanical Testing at LANSCE

Nuclear fuel encounters severe thermomechanical environments in which its mechanical response is determined by its microstructure, temperature and stress level histories. Simulating the response of such microstructures is crucial for predicting both performance and transient fuel mechanical responses and experimental verification of such predictions is therefore of great interest. While most of the deformation in a nuclear fuel rod occurs in the cladding, deformation of the fuel itself is still of interest with deformation mechanisms at operating temperature and above including creep, swelling, cracking as well as pellet-clad interaction. Characterization of these properties and understanding of the underlying deformation phenomena at operating or excursion temperatures is therefore of great importance for development and ultimately licensing of improved and novel nuclear fuel forms. Diffraction techniques offer unique insight on the atomistic (e.g. crystal structure) and microstructure (e.g. phase transformations, texture, defects) length scales and have a long history of providing unique data to inform relevant deformation models that enable the required predictive capabilities. For example, dislocations lead to diffraction peak broadening that can be characterized to estimate the dislocation density and study the role of dislocations on the deformation while measuring lattice strains allows to studie load sharing in two phase materials. In this report the requirements for a sample environment for high temperature deformation of nuclear fuels are defined. The HIPPO neutron time-of-flight diffractometer at LANSCE will host this sample environment and is also described. This instrument covers diffraction angles from 140° to 40° and is also equipped with an event-mode neutron imaging detector system, enabling energy-resolved neutron imaging in parallel with the diffraction that could measure sample temperature from Doppler broadening of neutron absorption resonances or measure pore densities from changes in the attenuation. Designs of devices to characterize thermomechanical properties of nuclear fuel without diffraction are also considered to guide the design. While this report is focused on applications for nuclear fuels, the device can also characterize cladding, moderator or structural materials and therefore contribute to other fields of research and development for advanced reactors. The temperatures planned to be reached are above 2000℃, thus enabling characterization of LWR reactor fuels under accident scenarios but also reaching temperatures of fuels developed for nuclear thermal propulsion and providing opportunities to characterize those. In conjunction with the energy-resolved neutron imaging detector, this setup would allow to measure neutron cross-sections at high temperatures, filling a gap towards development of reactors operating at high temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Rethinking PCMI Failures Under Transient Conditions

Under PCMI conditions, the cladding strain is driven by either pellet swelling during normal operation or pellet thermal expansion in transients. The loading in PCMI is displacement controlled because friction between the pellet and the cladding results in a strain compatibility at the pellet/cladding interface. Failure is known to occur under PCMI loading conditions during operation when normal power changes result in release of iodine (and other fission products) causing stress corrosion cracking where the crack-propagation rate is dependent upon the cladding stress but slow relative to strictly mechanical failures. Failures after normal power changes are often referred to as failure by pellet-cladding interaction or PCI to distinguish it from PCMI. Such failures are considered displacement-controlled failures, as noted above, and occur with limited plastic strain. This is different to what is typically thought of as PCMI failure under transient conditions, either Anticipated Operational Occurrences (AOOs) or Reactivity Insertion Accidents (RIAs), which is generally assumed to occur once the cladding reaches a mechanical strain limit. But there are two features in the typical PCMI failure in an RIA that are not consistent with the PCMI assumptions. The first is the ductile shear failure feature in cross section and the second is a large crack opening post failure (e.g., see OI-11 failure in NSRR program). The ductile feature is indicative of a plastic instability that requires very large local strains, and the large crack opening must be the result of a release of energy, both of which are not expected if stress in the cladding was only driven by thermal expansion in the pellet. It is concluded that there must be a mechanism (likely related to rod internal pressure) that eliminates strain compatibility between the pellet and cladding – allowing strain localization to occur in what are known as PCMI transient failures. In some sense, these PCMI failures may actually be early burst failures that occur when thermal expansion causes uniform strains greater than a defined limit and fission gas pressure eliminate the strain compatibility

Cantonwine, Paul↗
DOE OSTI2024

ANS Winter 2024 Summary: Optimizing the ATF-2Ramp Power Profile

When the Halden Boiling Water Reactor closed down in 2018, a need to restore the capability for in-reactor power ramp testing arose. Such testing is valuable for studying pellet-clad interaction phenomena in nuclear fuels. The data from these studies is of great interest to a number of research programs, including the accident-tolerant fuel (ATF) program at Idaho National Laboratory (INL). In 2022, Woolstenhulme et al. proposed several power ramp testing ideas using facilities at INL, including irradiation in the Transient Reactor Test Facility (better known as TREAT) and the Advanced Test Reactor (ATR) [1]. Worrall et al. [2] and Labossiere-Hickman et al. [3] subsequently performed feasibility studies for the ATR testing options in 2023. This summary further investigates the three-pin trefoil design (Fig. 1) for the proposed ATF-2Ramp Experiment discussed in Labossiere-Hickman et al. [3]. ATF-2Ramp is designed to operate in the center flux trap (CFT) of the ATR during a powered axial locator mechanism (PALM) cycle: a short, variable-powered cycle with an asymmetric power distribution. Previously, it was shown that tailoring the thickness of the hafnium (Hf) neutron shields (“mini-shrouds”) surrounding each pin offered a degree of control sufficient to achieve the programmatic linear heat generation rate (LHGR) targets for ATF-2Ramp during the high-power period of a PALM cycle. New work involves shortening the experiment test train for consistency with the fuel pins in ATF-2D [4] and then shaping the axial power profile of the three test pins.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Mechanical failure of fresh nuclear grade iron–chromium–aluminum (FeCrAl) cladding under simulated hot zero power reactivity initiated accident conditions

The reactivity initiated accident (RIA) is a postulated accident in light water reactors initiated by a rapid increase in reactivity which causes an increase in fission rate and fuel temperature. One potential mode of fuel system failure during RIA is pellet cladding mechanical interaction (PCMI) due to the rapid thermal expansion of the fuel pellet. Simulated PCMI experiments were performed by rapidly pressurizing iron-chromium-aluminum (FeCrAl) cladding tube samples using a hydraulic modified burst test system, causing the specimens to burst under a biaxial loading path. Deformation and rupture of the specimens were tracked with a telecentric lens and high-speed camera system. Outer surface strains were calculated using digital image correlation (DIC) on speckle patterns painted on the specimens’ outer surfaces. Experiments were conducted at approximately 275 °C, representative of hot zero power RIA conditions. The failure hoop strain was DIC-calculated between approximately 1.8–3.4%, corresponding to quasi-static energy depositions of approximately 110–260 calories per gram UO 2 assuming initial pellet-cladding contact. These conditions are very similar to the proposed energy deposition limit of 150 calories per gram UO 2 for unirradiated zirconium-based cladding in US Nuclear Regulatory Commission (NRC) Draft Regulatory Guide 1327. A very small strain rate dependence was observed in the data, with the magnitude of the failure hoop strain decreasing slightly with increasing strain rate. This observed dependence may be relevant because the FeCrAl cladding strain rate will be approximately 20% higher than in zirconium-based cladding due to the harder neutron spectrum and resultant shorter neutron generation time.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Status Update on the SCIP-IV Program in 2020

Phase 1 of the Studsvik Cladding Integrity Project (SCIP) was launched in 2004. It was a 5-year Organisation for Economic Co-operation and Development (OECD) / Nuclear Energy Agency (NEA) joint project operated by Studsvik with about 30 participating organizations, including regulatory bodies, research institutions, utilities, and fuel suppliers from 13 different countries. The initial SCIP program focus was directed toward studying basic phenomena of fuel rod failures driven by pellet cladding mechanical interaction to get a better understanding of fundamental failure mechanisms. SCIP IV is a continuation of the SCIP programs and is planned to be another five-year project that started in July 2019. It is organized and managed in a manner similar to the earlier SCIP programs, including both a technical group and a management board. The work of SCIP-IV includes significant research into issues related to the back end of the fuel cycle, including spent nuclear fuel (SNF) storage and transportation issues.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗