Irradiation-induced Creep Research Articles

Detection of dislocation motion in atomistic simulations of nanocrystalline materials

A method for identifying dislocation motion in atomistic simulations is presented. While identifying static and moving dislocations within a single crystal or a combination of such is well established, the method described here is tailored to identify dislocation motion by correlating the displacements of individual atoms. This facilitates the identification of dislocation motion in complex structural arrangements, and allows the specific contribution to plastic deformation, due to dislocation motion, to be separated from that of other mechanisms. The method is applied to test cases in crystals and grain boundaries (GBs), in which irradiation-induced creep (IIC) was induced. It is shown that the method singles out the moving dislocations from among the dislocation forest at GBs, thus identifying the specific reactions driving the distortion at any given time. This enables the study of dislocation processes in the presence of realistic obstacles, and the study of the effects of microstructure on dislocation mobility. As an example of such a study, the method is applied to rule out intragranular slip, and to estimate the contribution of dislocation motion to strain, in a NC undergoing IIC.

Fuel Performance Analysis of Fast Flux Test Facility MFF-3 and -5 Fuel Pins Using BISON with Post Irradiation Examination Data

Using the BISON fuel-performance code, simulations were conducted of an automated process to read initial and operating conditions from the Pacific Northwest National Laboratory (PNNL) database and reports, which contain metallic-fuel data from the Fast Flux Test Facility (FFTF) MFF Experiments. This work builds on previous modeling efforts involving 1977 EBR-II metallic fuel pins from experiments. Coupling the FFTF PNNL reports to BISON allowed for all 338 pins from MFF-3 and MFF-5 campaigns to be simulated. Each BISON simulation contains unique power and flux histories, axial power and flux profiles, and coolant-channel flow rates. Fission-gas release (FGR), fuel axial swelling, cladding profilometry, and burnup were all simulated in BISON and compared to available post-irradiation examination (PIE) data. Cladding profilometry, FGR, and fuel axial swelling simulation results for full-length MFF metallic pins were found to be in agreement with PIE measurements using FFTF physics and models used previously for EBR-II simulations. The main two peaks observed within the cladding profilometry were able to be simulated, with fuel-cladding mechanical interaction (FCMI), fuel-cladding chemical interaction (FCCI), and thermal and irradiation-induced creep being the cause. A U-Pu-Zr hot-pressing model was included in this work to allow pore collapse within the fuel matrix. This allowed better agreement between BISON-simulated cladding profilometry and PIE measurements for the peak caused by FCMI. This work shows that metallic fuel models used to accurately represent fuel performance for smaller EBR-II pins may be used for full-length metallic fuel, such as FFTF MFF assemblies and the Versatile Test Reactor (VTR). As new material models and PIE measurements become available, FFTF MFF assessment cases will be reassessed to further BISON model development.

Irradiation creep measurement and microstructural analysis of chromium nitride–coated zirconium alloy using pressurized tubes

Environmental barrier coatings for Zr-based materials are currently under development to reduce oxidation and embrittlement in light-water reactors. Chromium nitride is one such candidate for this application, particularly as accident-tolerant fuel cladding. However, quantifying the impact of coatings on the irradiation-induced creep of zircaloy (Zry) is critical as this mechanism often exceeds thermal creep rates under light-water reactor operating conditions and can be a limiting design characteristic. Additionally, examining irradiation effects in the microstructure at the coating interface is key to understanding the compatibility of the material system. To accelerate the experimental measurement of irradiation creep and microstructure evolution in CrN-Zry, compact, pressurized creep tubes were fabricated and irradiated in the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Miniature, thin-walled rodlets fabricated from annealed Zr-Sn barstock were coated with CrN using physical vapor deposition (PVD) to nominal thicknesses of 4 and 8 μm. Coated and uncoated rodlet specimens were internally pressurized and welded, generating nominal circumferential hoop stresses of 0, 90, or 180 MPa under 300∘C irradiation conditions. Twelve specimens were measured diametrically prior to irradiation using a low-cost, automated, contactless laser profilometer developed for this work. Specimens were irradiated in sealed capsules for one 25-day HFIR cycle, accumulating approximately 1.8×1021n/cm2 fast fluence (En>1.0MeV). The irradiated samples were retrieved and remeasured using the same profilometry system in a shielded hot cell facility. Irradiation creep between specimens was compared using standard statistical tests and showed that both thicknesses of CrN coating had a negligible effect on the irradiation creep strain of the Zry material. Microstructure characterization of pre- and post-irradiated CrN-Zry specimens showed minimal changes due to irradiation but did show a substantial O-rich region at the Zry-CrN interface.

Investigation on the Irradiation-thermal–mechanical coupling behaviors of the solid Microencapsulated fuel rod

Solid Microencapsulated Fuel (SMF) is a kind of new accident-tolerant fuel (ATF), where the TRISO coated fuel particles are dispersed directly into a type of matrix in the form of a solid rod. SMF is expected to provide improved performance because of the elimination of cladding tube and associated failure mechanisms. This study aims to investigate the irradiation-thermal–mechanical coupling behavior of SMF rod under normal operation of the 200 MW Nuclear Heating Reactor (NHR200-II) using finite element method (FEM). Two kinds of SMFs have been evaluated and compared -with the fuel particles dispersed into a silicon carbide matrix or a zirconium matrix. The finite element models, considering comprehensive coupling behavior (including irradiation-induced dimensional change, irradiation-induced creep, thermal expansion, fission gas pressure, heat generation and conduction, elasticity and plasticity, etc.), were developed based on the ABAQUS platform and accurately verified. The study also investigated the impacts of key parameters on the coupling behavior of SMF, including particle kernel diameter, non-fuel zone width, extent of particle aggregation, and operational conditions. The results revealed that SMFs exhibit excellent heat transfer capability, resistance to deformation, and a stress-relaxed state under normal operation. Specifically, SMF-Zr exhibited a higher centerline temperature but lower levels of deformation and stress compared to SMF-SiC. The effects of fuel kernel diameter and non-fuel zone width were found to be limited, whereas higher particle aggregation and harsher operational conditions indeed increase temperature, deformation, and stress, particularly during the initial irradiation stage. Moreover, a higher creep coefficient of pyrolytic carbon leads to a more relaxed stress state. This study lays a fundamental basis for the design and optimization of this kind of solid particle-dispersed fuel.

Irradiation-Induced Creep and Re-Sintering of Large Grain Sized UO2 Fuel

Irradiation-Induced Creep and Re-Sintering of Large Grain Sized UO2 Fuel

Investigation on effect of neutron irradiation on welding residual stresses in core shroud of pressurized water reactor

This paper presents the results of investigating the change in welding residual stresses of the core shroud, which is one of subcomponents in reactor vessel internals, performing finite element analysis. First, the welding residual stresses of the core shroud were calculated by applying the heat conduction based lumped pass technique and finite element elastic-plastic stress analysis. Second, the temperature distribution of the core shroud during the normal operation was calculated by performing finite element temperature analysis considering gamma heating. Third, through the finite element viscoelastic-plastic stress analysis using the calculated temperature distribution and setting the calculated residual stresses as the initial stress state, the variation of the welding residual stresses was derived according to repeating the normal operation. In the viscoelastic-plastic stress analysis, the effects of neutron irradiation on mechanical properties during the cyclic normal operations were considered by using the previously developed user subroutines for the irradiation agings such as irradiation hardening/embrittlement, irradiation-induced creep, and void swelling. Finally, the effect of neutron irradiation on the welding residual stresses was analysed for each irradiation aging. As a result, it is found that as the normal operation is repeated, the welding residual stresses decrease and show insignificant magnitudes after the 10th refueling cycle. In addition, the irradiation-induced creep/void swelling has significant mitigation effect on the residual stresses whereas the irradiation hardening/embrittlement has no effect on those.

X-ray characterization of anisotropic defect formation in SiC under irradiation with applied stress

High-energy x-ray diffraction is employed to gain insight into the irradiation-induced creep behavior of silicon carbide (SiC). Polycrystalline β-SiC specimens were simultaneously exposed to elevated temperature neutron-irradiation and mechanically applied stresses. The structural disordering was subsequently examined using two-dimensional x-ray diffraction. The intensity of the (111) shoulder peak, an indication of stacking disorder, increased when the specimens were irradiated under tensile stress. This is the first observation of nanoscale stress-induced stacking disorder in SiC at low neutron fluences. These findings suggest stress-induced preferential nucleation and/or growth of defect clusters as a key creep mechanism in neutron irradiated SiC.

Improved Computational Fluid Dynamics Framework for Reactor Core Baffle Swelling Assessment

Abstract This paper presents an improved estimation of reactor core baffle temperature distribution, during operation, at the nominal power level to address swelling problems of the reactor internals. Swelling is the main limiting factor in the reactor core internals long term operation of VVER-1000 nuclear units. The material irradiation-induced swelling and creep models are very sensitive to temperature distribution in metal; thus, a more detailed analysis of the core baffle metal thermohydraulic cooling characteristics is required. A framework for the computational fluid dynamics (CFD) analysis of VVER-1000 reactor baffle cooling is presented. First, an analytical model was developed to obtain boundary conditions (BCs) and simplify CFD analysis. Second, the CFD analysis was performed using 60 deg symmetry, which included core, baffle, and core barrel, and it is limited by the height of the baffle. Core is simplified as an equivalent coolant domain with considering of spatial volumetric energy release. Core baffle is presented as monolithic body with considering of gamma-ray heat generation. Model includes cooling ribs and simplified geometry of connecting studs, with cooling flow of the coolant through the nuts grooves. Calculated convection coefficient and temperature are in good agreement with analytical model and give a more accurate result comparing to RELAP5/mod3.2. Obtained temperature field was used to estimate baffle swelling process and justify safe long term operation of the reactor internals.

Hypothesis for a mechanism of beam-induced motion in cryo-electron microscopy.

Estimates of heat-transfer rates during plunge-cooling and the patterns of ice observed in cryo-EM samples indicate that the grid bars cool much more slowly than do the support foil and sample near the middle of the grid openings. The resulting transient temperature differences generate transient tensile stresses in the support foil. Most of this foil stress develops while the sample is liquid and cooling toward its glass transition T g, and so does not generate tensile sample stress. As the grid bars continue cooling towards the cryogen temperature and contracting, the tensile stress in the foil is released, placing the sample in compressive stress. Radiation-induced creep in the presence of this compressive stress should generate a doming of the sample in the foil openings, as is observed experimentally. Crude estimates of the magnitude of the doming that may be generated by this mechanism are consistent with observation. Several approaches to reducing beam-induced motion are discussed.

A Mathematical Model of Radiation-Induced Shape Change in Fuel Subassemblies of the BN-Type Reactor Core and Its Implementation in the ANSYS Software Package

One of the main operability criteria for fuel subassemblies (FSAs) in sodium-cooled fast reactor cores, i.e., the criterion of tolerable shape change in the hexagonal wrapper tube, is formulated. The equations that enable one to investigate the kinetics of the stress-strain state of a three-dimensional body have been adapted to operating conditions of the FSAs. A mathematical model of radiation-induced shape change in ferritic-martensitic EP-450 grade steel is proposed. With regard to the proposed model and data on the radiation-induced shape change in other currently used and prospective BN reactor core structural materials, blocks for recording radiation-induced swelling and radiation-induced creep were developed for the ANSYS software package, which made it possible to utilize its potential in this area. The test case with the proposed models of the radiation-induced swelling and creep demonstrates that the developed blocks describe sufficiently well the radiation-induced shape change in the examined structural materials exposed to radiation. Calculation of the radiation-induced shape change in the FSA hexagonal wrapper tube has been performed at various radiation-induced swelling rates and radiation-induced creep moduli. The calculated results and the results of post-irradiation inspection of the FSA dimensions are compared. Recommendations for use of the proposed models designed to calculate and estimate the radiation-induced shape change and define the stress-strain state of the FSAs are made.

Irradiation induced creep in nanocrystalline high entropy alloys

Irradiation induced creep (IIC) compliance in NiCoFeCrMn high entropy alloys is measured as a function of grain size (30 < x < 80 nm) and temperature (23–500 °C). For 2.6 MeV Ag3+ irradiation at a dose rate of 1.5×10–3 dpa−1s−1 the transition from the recombination to sink limited regimes occurs at ∼ 100 °C. In the sink-limited regime, the IIC compliance scales inversely with grain size, consistent with a recently proposed model for grain boundary IIC. The thermal creep rate is also measured; it does not become comparable to the IIC rate, however, until ∼ 650 °C. The results are discussed in context of defect kinetics in irradiated HEA systems.

In-pile tensile creep of chemical vapor deposited silicon carbide at 300 °C

Irradiation-induced creep is one of the key material properties considered in designing structural components for nuclear reactors. This paper presents results for in situ irradiation-induced creep of chemical vapor–deposited 3C silicon carbide studied by instrumented irradiation in the Halden reactor in Norway. The specimens examined were irradiated at 300 °C and up to 2.5 × 1024 n/m2 (E > 0.1 MeV) under uniaxial tensile stress of <5 or 100 MPa. Irradiation-induced creep strain was defined as the differential time-dependent strain between the two specimens. Based on the dimensional inspections before and after irradiation, an axial primary creep strain of 0.06% was obtained at the end of irradiation. The lattice constant precisely determined from high-energy x-ray diffraction analysis showed a lattice expansion roughly accounting for the primary irradiation creep strain. Analysis of data from this and previous studies indicates that creep strain is significantly dependent on at least one of the experimental conditions, such as loading mode, neutron spectrum/flux, and material grade.

Mathematical model of radiation-induced shape change in fuel subassemblies of the BN-type reactor core and its implementation in ANSYS software package

One of main operability criteria for fuel subassemblies (FSAs) in fast sodium reactor cores, i.e. the criterion of tolerable shape change in hexagonal wrapper tube is formulated. The equations which enable one to inquire into kinetics of the stress-strained state of a three-dimensional body are adapted to FSAs operating conditions. A mathematical model of radiation-induced shape change in ferritic-martensitic steel of grade EP-450 is proposed. With regard to the proposed model and data on the radiation-induced shape change in other current and prospective structural materials of BN reactor cores, blocks for recording of radiation-induced swelling and radiation-induced creep are developed for ANSYS software package, which made it possible to utilize its potentials within this area of focus. The performed test case with proposed models of the radiation-induced swelling and radiation-induced creep demonstrates that the developed blocks sufficiently describe the radiation-induced shape change in the examined structural materials exposed to radiation. A calculation of the radiation-induced shape change in the FSA hexagonal wrapper tube with various speeds of the radiation-induced swelling and the radiation-induced creep moduli is performed. The calculations results and the results of FSAs post-irradiation dimension inspection are compared. Recommendations for use of the proposed models aimed at performing calculations, as well as estimating the radiation-induced shape change and defining the stress-strained state of FSAs are made.

A model of gas-induced effective expansion strain for porous carbon materials in irradiation environments

The application of fully ceramic microencapsulated (FCM) fuels to light water reactors has attracted increasing attention. As the main storage space for fission gases and some other gases in FCM fuels, the microstructure and thickness of porous carbon buffer layer should be optimally designed. In this study, a model of gas-induced effective expansion strain rate for porous carbon materials is developed based on meso-mechanical method and homogenization theory. By combining theoretical analysis with finite element (FE) simulation, the resulting model is verified, which considers the irradiation-induced creep deformation of dense PyC skeleton, the gas contained in the pores of the buffer layer, the accumulated effective volumetric strain and the external pressure. The effects of temperature, fission rate, kernel materials, initial porosity, external hydrostatic pressure and creep coefficient of dense PyC skeleton are investigated. This work is expected to provide a theoretical reference for optimization design of a porous carbon buffer layer, and the developed model can be used for further simulation of irradiation-induced thermo-mechanical behavior in FCM fuels.

Historical experiment to measure irradiation-induced creep of graphite

This paper presents historical results of graphite irradiation-induced creep experiments that were performed at Oak Ridge National Laboratory from the 1950's to the 1970's. These experiments were performed at temperatures from 150°C to 1000 °C, and bend stresses ranging from 500 to 5000 psi (∼3.3–34.5 MPa). The experimental setup utilized in-situ measurement of specimen displacement, on-line applied stress control, and the ability to change stress during the experiment. The different stress conditions showed that the primary creep strain and the steady-state creep rates both have a linear stress dependence. The temperature range used in this work resulted in trends that have not be previously presented in the literature: 1) a linear dependence of primary creep strain on temperature, and 2) the shape of steady state creep rate versus temperature (see graphical abstract). The maximum dose in the specimens was 0.9 dpa, which is sufficient to achieve steady-state creep without the structural changes that alter the observed creep behavior. The results from this experiment provide evidence that dispels that the pinning-unpinning model describes the mechanism of irradiation creep in graphite. Instead these results suggest a dislocation climb mechanism is the probable mechanism for creep within the crystalline regions.

Effect of Stress on Irradiation-induced Creep and Swelling of Fe-18Cr-10Ni-Ti Steel Pressurized Specimens Irradiated in the BOR-60 Reactor

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High temperature irradiation induced creep in Ag nanopillars measured via in situ transmission electron microscopy

Irradiation induced creep (IIC) rates are measured in compression on Ag nanopillar (square) beams in the sink-limited regime. The IIC rate increases linearly with stress at lower stresses, i.e. below ≈2/3 the high temperature yield stress and parabolically with pillar width, L, for L less than ≈300 nm. The data are obtained by combining in situ transmission electron imaging with simultaneous ion irradiation, laser heating, and nanopillar compression. Results in the larger width regime are consistent with prior literature.

Finite element modelling of multilayer Advanced Gas-cooled Reactor bricks and creep interaction

The structural integrity of nuclear graphite bricks is important in Advanced Gas-cooled Reactors (AGRs) as they not only provide moderation but channels for fuel, cooling and control rods. AGR graphite moderator bricks are subjected to fast neutron irradiation and radiolytic oxidation during reactor operation, leading to component dimensional and material properties changes. With irradiation ageing the graphite components deform in the axial and radial directions causing changes to bore diameter and brick length. Furthermore, deformation at the brick end faces (called dishing) can lead to the formation of axial gaps between the bricks which can then lead to fuel channel bowing. However, it is believed that irradiation creep reduces the potential size of these axial gaps and retains bricks in contact to some extent. Therefore, as the reactors age it is important to understand the nature of this contact behaviour between bricks in the fuel channels. This paper focuses primarily on the finite element modelling of contact behaviour between bricks and the effects of irradiation-induced creep on contact conditions and hence, ovality of the brick bore. Multilayer fuel brick models and a brick with a rigid body on the top surface have been modelled. The results show that multilayer models are required to understand the contact conditions between the bricks throughout the life of a reactor.

Surface morphology and microstructure evolution of IG-110 graphite after xenon ion irradiation and subsequent annealing

IG-110 graphite samples were polished and irradiated with Xe ions at various fluences, then annealed at high temperatures up to 1100 °C. After irradiation, small hills were found on the polished surfaces, indicating an anisotropic swelling induced by irradiation. Around 30% swelling at a fluence of 2 × 1015 ions/cm2 was characterized using atomic force microscopy. Severe swelling of the graphite crystallites caused stresses between adjacent crystallites, but leaved no intergranular cracks on the polished surface, which was ascribed to irradiation-induced creep of graphite. The pore morphology was affected by the anisotropic swelling. We found many contracted pores but only one expanded pore, indicating a decreased porosity induced by irradiation. After annealing at 1100 °C, TEM characterization showed clearly increased lattice order and decreased width of the (002) diffraction arc, indicating the annihilation of dislocations and recovery of basal plane rotations. Annealing-induced recrystallization of damaged graphite led to recovery of the crystallites' swelling and many small cracks appearing on the samples' surfaces.

Irradiation-induced creep in metallic nanolaminates characterized by In situ TEM pillar nanocompression

This work reports on irradiation-induced creep (IIC) measured on nanolaminate (Cu-W and Ni-Ag) and nanocrystalline alloys (Cu-W) at room temperature using a combination of heavy ion irradiation and nanopillar compression performed concurrently in situ in a transmission electron microscope. Appreciable IIC is observed in multilayers with 50 nm layer thicknesses at high stress, ≈½ the yield strength, but not in multilayers with only 5 nm layer thicknesses.