Irradiation Creep Research Articles

Skeleton-creep based bubble growth model and multi-scale mechanical constitutive model for U-10Mo fuels under irradiation

The macroscale irradiation swelling deformations of nuclear fuels induced by the solid and gas fission products are one of the most concerned problems in nuclear reactor core design. Based on continuum mechanics, an innovative macroscale volume growth strain model is proposed for U-10Mo fuels, with the coupling contributions of irradiation creep, fission solid swelling and thermal expansion of fuel skeleton. Correspondingly, new three-dimensional mechanical constitutive models and stress update algorithms are also developed for the multi-scale irradiation-deformation and stress analysis. Good agreements of the predictions with diverse experimental data validate the effectiveness of newly developed models. It is obtained that the growth of fission gas bubbles is dominated by the creep deformations of the surrounding fuel skeleton, depending on the external hydrostatic pressure and the internal bubble pressure related to the intra-granular fission gas atom diffusion; the influences of temperature, external hydrostatic pressure and fuel-skeleton creep rate coefficient are found to become prominent at higher burnup levels, dominantly induced by the coupling of fuel-skeleton creep and local porosity. This study provides a new insight into the mechanism of bubble growth in the solid skeleton, revealing the coupling of irradiation swelling with fuel-skeleton creep.

Multiphysics simulation of VVER-1200 fuel performance during normal operating conditions

Nuclear fuel performance modeling and simulation are critical tasks for nuclear fuel design optimization and safety analysis under normal and transient conditions. Fuel performance is a complicated phenomenon that involves thermal, mechanical, and irradiation mechanisms and requires special multiphysics modules. In this study, a fuel performance model was developed using the COMSOL Multiphysics platform. The modeling was performed for a 2D axis-symmetric geometry of a UO2 fuel pellet in the E110 clad for VVER-1200 fuel. The modeling considers all relevant phenomena, including heat generation and conduction, gap heat transfer, elastic strain, mechanical contact, thermal expansion, grain growth, densification, fission gas generation and release, fission product swelling, gap/plenum pressure, and cladding thermal and irradiation creep. The model was validated using a code-to-code evaluation of the fuel pellet centerline and surface temperatures in the case of constant power, in addition to validation of fission gas release (FGR) predictions. This prediction proved that the model could perform according to previously published VVER nuclear fuel performance parameters. A sensitivity study was also conducted to assess the effects of uncertainty on some of the model parameters. The model was then used to predict the VVER-1200 fuel performance parameters as a function of burnup, including the temperature profiles, gap width, fission gas release, and plenum pressure. A compilation of related material and thermomechanical models was conducted and included in the modeling to allow the user to investigate different material/performance models. Although the model was developed for normal operating conditions, it can be modified to include off-normal operating conditions.

Accelerated fission rate irradiation design, pre-irradiation characterization, and adaptation of conventional PIE methods for U-10Mo and U-17Mo

Metallic U alloys have high U density and thermal conductivity and thus have been explored since the beginning of nuclear power research. Alloys of U with modest amounts of Mo, such as U-10 wt % Mo (U-10Mo), are of particular interest because the γ-U crystal structure in this alloying addition shows prolonged stability in reactor service. Historically, radiation data on U-10Mo fuels were collected in Na fast reactors or lower temperature research reactor conditions, but little is known about irradiation behavior, particularly swelling and creep, at irradiation temperatures between 250 and 500°C. This work discusses the methodology and pre-irradiation characterization results from a U-Mo irradiation campaign performed in the High Flux Isotope Reactor at Oak Ridge National Laboratory. U-10Mo and U-17Mo samples irradiations are being completed at temperatures ranging from 250 to 500°C to three targeted fission densities between 2 × 1020 and 1.5 × 1021 fissions per cubic centimeter. Swelling measurement of the specimen sizes studied here required development and assessment of new methods for volume determination before and after irradiation. Laser profilometry and X-ray computation tomography (XCT) were used to provide preirradiation characterization of samples to determine the error and applicability of each to determine swelling following irradiation. These outcomes are contextualized through use of BISON simulations performed to assess the predicted expansion of U-Mo fuels subjected to the irradiation conditions of this work. Use of existing BISON fuel performance models predicted a maximum of 7% swelling under the irradiation conditions of this study. Pre-irradiation characterization revealed the as-cast U-Mo fuel samples were uniformly large-grained fully cubic U crystals with small U-C/N bearing precipitates and pores distributed throughout. Samples were found to contain a bulk porosity between .4 and 3% because of the casting process. Local porosity in areas far from large, interconnected pores was found by Slice-and-View to be under .2%. Nanometer-sized precipitates rich in C and N were identified in all samples, likely because of impurities during the fabrication process. Dendritic bands were also observed throughout the samples. These bands were characterized by variable Mo content that deviated from the overall Mo content by 2–3 wt %. No other microstructural features were correlated to these bands. Mechanical properties were found to be slightly strengthened compared to literature reports of bulk U-Mo fuels due to the nano-scale precipitates throughout the sample.

Analysis of Radiation Creep Problems Considering Pore Growth Ductile Fracture by Rice–Tracey–Huang Models and Kachanov’s Solution for a Spherical Cavity. Part 2. Validity Check of the Governing Equations

Analysis of Radiation Creep Problems Considering Pore Growth Ductile Fracture by Rice–Tracey–Huang Models and Kachanov’s Solution for a Spherical Cavity. Part 2. Validity Check of the Governing Equations

Physics-based model of irradiation creep for ferritic materials under fusion energy operation conditions

Irradiation creep is known to be an important process for structural materials in nuclear environments, potentially leading to creep failure at temperatures where thermal creep is generally negligible. While there is a great deal of data for irradiation creep in steels and zirconium alloys in light water reactor conditions, much less is known for first wall materials under fusion energy conditions. Lacking suitable fusion neutron sources for detailed experimentation, modeling, and simulation can help bridge the dose-rate and spectral-effects gap and produce quantifiable expectations for creep deformation of first wall materials under standard fusion conditions. In this paper, we develop a comprehensive model for irradiation creep created from merging a crystal plasticity representation of the dislocation microstructure and a defect evolution simulator that accounts for the entire cluster dimensionality space. Both approaches are linked by way of a climb velocity that captures dislocation-biased defect absorption and a dislocation strengthening term that reflects the accumulation of defect clusters in the system. We carry out our study in Fe under first wall fusion reactor conditions, characterized by a fusion neutron spectrum with average recoil energies of 20 keV and a damage dose rate of ≈3×10−7 dpa/s at temperatures between 300 and 800 K.

In-situ irradiation creep behavior of dilute Al-Sc alloys: Using a thin film bulge test method

• In situ thin film bulge test on nc-AlSc is effective method for irradiation induced creep. • Preexisting Al 3 Sc precipitates suppress irradiation induced creep deformation. • Transition regime of AlSc binary system is located between room temperature and 75 °C. In-situ irradiation induced creep(IIC) measurements on a nanocrystalline AlSc 1.1 binary alloy were performed using a thin-film bulge test technique. A thin film bulge technique was employed since any substrate effect can be excluded, and hence is beneficial to analyze the intrinsic properties of specimen. Here, an as-deposited film and a pre-annealed film, both underwent heavy ion irradition with 1.8 MeV Kr ions. Al 3 Sc precipitates in the pre-annealed samples appear to reduce the irradiation creep compliance. Dissolution of these precipitates during irradiation led to increased irradiation creep rates. Microstructural evolution during irradiation was investigated by transmission electron microscopy and high resolution transmission electron microscopy.

Analysis of Radiation Creep Problems Considering Pore Growth Ductile Fracture by Rice–Tracey–Huang Models and Kachanov’s Solution for a Spherical Cavity. Part 1. Initial Preconditions and Models for the Formulation of the Governing Equations

Analysis of Radiation Creep Problems Considering Pore Growth Ductile Fracture by Rice–Tracey–Huang Models and Kachanov’s Solution for a Spherical Cavity. Part 1. Initial Preconditions and Models for the Formulation of the Governing Equations

The effect of cold work on irradiation creep of Zircaloy-2 alloys at low temperature

Internally pressurized capsules made from Zircaloy-2 tubing were irradiated in the National Research Universal (NRU) reactor at Chalk River Laboratories (CRL) at a temperature of approximately 57°C (330 K) in a fast neutron flux of about 3.37 × 1017 n·m−2·s−1 (E>1 MeV). The purpose of this experiment was to study the effect of cold work and crystallographic texture on biaxial stress irradiation creep. The creep capsules were cold worked in both transverse and axial directions. The applied hoop stress during irradiation ranged from 0 to approximately 200 MPa. Two irradiation cycles were completed with a fast neutron fluence of 3.18 × 1025 n·m−2 (E >1 MeV). Diametral and axial strains were measured after each irradiation cycle.This experiment has provided the following insights on Zircaloy-2 alloys irradiated at low temperature. The axial strain increases with fast neutron fluence, while the diametral strain either increases or decreases with fluence depending on the value of hoop stress and the amount of cold work. Both the diametral and axial strain rates increase non-linearly with the hoop stress. There is a weak dependence of the strain rate on cold work at low hoop stresses (50-150 MPa) suggesting that dislocation slip is not a dominant strain producing mechanism at irradiation temperature of about 57°C. The non-linear dependence of the strain rate on cold work that is observed at 200 MPa indicates that dislocation slip mechanisms start to dominate at this stress level.Dislocation densities of the creep capsules were measured by X-ray line broadening, and their structures were further analyzed using transmission electron microscopy (TEM). From these results, the weak dependence of the creep rate on cold work at low stresses can be explained using rate theory model formalism.

Prediction of the in-reactor deformation of Zr-2.5wt%Nb pressure tubes using the crystal plasticity finite element method framework

To predict the irradiation induced deformation of the pressure tube material in Canada Deuterium Uranium (CANDU) reactors, a deformation model based on the crystal plasticity finite element method framework was developed in this study. Based on the principle of superimposition, three deformation mechanisms—thermal creep, irradiation growth, and irradiation creep—were incorporated into the model. The model parameters were optimized and verified using in-service inspection (ISI) data from Korean CANDU reactors by considering the irradiation effect on deformation mechanisms and the influence of incompatible grain interactions. We found that the model with optimized parameters could successfully describe the results of not only the pressure tube ISI data but also the simulated mini capsules experiments in a test reactor. By analyzing the optimized irradiation creep parameters and results, the possibility of diffusion-base mechanisms was elucidated and the anisotropy in the activation energy terms along the tube directions was confirmed. For predictions using the model, three strategies for extracting the optimized parameters from available measurement data were examined. As a result, we found that, the better prediction ability can be achieved when only the latest data measured at the same reactor before the predictions are used to optimize the model parameters (method 3).

Effect of manufacturing route on thermal creep behaviour of Zr-2.5Nb pressure tube alloy used in Indian PHWR

Indian Pressurized Heavy Water Reactors (IPHWRs) uses Zr-2.5Nb alloy pressure tube (PT) in cold work and stress relieved condition. These PTs are subjected to extreme conditions of stress, temperature and irradiation. Since past three decades of PHWR operation in India, fabrication routes of PTs for IPHWR has undergone several modifications to enhance its in-reactor performance. The deformation of PT in PHWR occurs due to irradiation creep, irradiation growth and thermal creep. In the present study, the different generations of PTs used in IPHWR are tested for its thermal creep behaviour. The accelerated creep rupture tests were carried out using flat tensile specimens with its axis parallel to the longitudinal direction of the PT in the temperature range of 350 to 450°C under stresses in the range of 0.7 to 0.9 times of yield strength at the respective temperatures. The minimum creep rates were evaluated along with the stress exponents and activation energies. The values of stress exponent and activation energy revealed the underlying creep mechanisms responsible for the creep deformation. Microstructural examination was carried out to understand the need and efficacy of the new fabrication route. The observed thermal creep behaviour of all the generations of pressure tubes were rationalised in terms of their respective texture and microstructures.

Irradiation damage concurrent challenges with RAFM and ODS steels for fusion reactor first-wall/blanket: a review

Reduced activation ferritic martensitic (RAFM) and oxide dispersion strengthened (ODS) steels are the most promising candidates for fusion first-wall/blanket (FW/B) structures. The performance of these steels will deteriorate during service due to neutron damage and transmutation-induced gases, such as helium/hydrogen, at elevated operating temperatures. Here, after highlighting the operating conditions of fusion reactor concepts and a brief overview, the main irradiation-induced degradation challenges associated with RAFM/ODS steels are discussed. Their long-term degradation scenarios such as (a) low-temperature hardening embrittlement (LTHE)—including dose-temperature dependent yield stress, tensile elongations, necking ductility, test temperature effect on hardening, Charpy impact ductile-to-brittle transition temperature and fracture toughness, (b) intermediate temperature cavity swelling, (c) the effect of helium on LTHE and cavity swelling, (d) irradiation creep and (e) tritium management issues are reviewed. The potential causes of LTHE are discussed, which highlights the need for advanced characterisation techniques. The mechanical properties, including the tensile/Charpy impact of RAFM and ODS steels, are compared to show that the current generation of ODS steels also suffers from LTHE, and shows irradiation hardening up to high temperatures of ∼400 °C–500 °C. To minimise this, future ODS steel development for FW/B-specific application should target materials with a lower Cr concentration (to minimise α′), and minimise other elements that could form embrittling phases under irradiation. RAFM steel-designing activities targeting improvements in creep and LTHE are reviewed. The need to better understand the synergistic effects of helium on the thermo-mechanical properties in the entire temperature range of FW/B is highlighted. Because fusion operating conditions will be complex, including stresses due to the magnetic field, primary loads like coolant pressure, secondary loads from thermal gradients, and due to spatial variation in damage levels and gas production rates, an experimentally validated multiscale modelling approach is suggested as a pathway to future reactor component designing such as for the fusion neutron science facility.

EFFECT OF IRRADIATION AND THERMO-INDUCED PROCESSES ON REACTOR IN VESSEL ELEMENTS DURING LONG-TERM OPERATION

A method for determining the effect of inhomogeneous temperature distribution, irradiation damage, irradiation swelling, irradiation creep of the material on the long-term strength of reactor in vessel internals and their deformation has been developed. The deformed state of the WWER type nuclear reactor baffle was determined. Constitutive equations taking into account the hidden damage of the material are proposed. According to the results of computer simulations, the strain fields of the section of the baffle during its operation for 60 years were analyzed, the place of possible occurrence of a macroscopic defect was determined.

An innovative multi-layer shell model for in-pile thermo-mechanical behavior analysis of plate-type fuel assemblies

In this study, an innovative large-deformation multi-layer shell model is developed. In order to improve the efficiency of simulating the thermo-mechanical coupling behavior in complex plate-type fuel assemblies, the fuel plates are simulated by 2D shell element, with the irradiation swelling and creep effects involved. The specific theoretical frame for the kinematic models and three-dimensional constitutive models is presented, together with the generalized force update algorithm. The application of the developed shell model in the thermal–mechanical behavior simulation of a monolithic U-Mo/Al fuel assembly indicates that (1) the out-of-plane displacements of the side plates can be precisely calculated, with the calculation time ∼92% shorter than that of a full 3D simulation of 308-day irradiation behavior; (2) the out-of-plane displacements of the fuel plates excluding the peripheral ones can also be well obtained, with small relative deviations to the full 3D simulation results; (3) using 3D solid models for the peripheral fuel plates, a fuel assembly finite element (FE) model with 2D&3D geometric models for the fuel plates is proposed with a total computation time decrease of 64%, and the precise of deformation simulation results can be improved comprehensively. This research provides important theoretical models and algorithms for the irradiation-induced deformation analysis of complex plate-type fuel assemblies.

Further development of the fission gas swelling model for U-10Mo fuels

The irradiation swelling of U-10Mo fuels is one of the critical factors to affect the in-pile structural integrity and dimensional stability of fuel plates. In this study, a mechanistic model of fission gas swelling for U-10Mo fuels is newly developed with a new evolution model of bubble density, considering the fission gas diffusion behavior and the grain recrystallization. The developed theoretical frame is validated by comparing the predictions in this study with diverse experimental data quantitatively and qualitatively, including the fission gas swelling, total irradiation swelling, the evolution rules of the bubble density and the bubble size, together with the thickness increments of fuel foil in a monolithic U-10Mo/Al fuel plates. The correlation of the irradiation creep with the fission-gas-swelling induced porosity for U-10Mo fuels is firstly considered in the simulation of irradiation-induced thermal-mechanical behaviors in Al-cladded monolithic fuel plates, and a new creep coefficient of dense U-10Mo fuels is identified as 180 × 10−22 mm3/MPa. The research results indicate that the developed fission gas swelling model has a satisfactory prediction ability, especially for higher burnup cases, and can describe the dominant dependences of external pressure and temperature found in the related experiments. In addition, parametric studies have been carried out to investigate the influences of temperature, external pressure and grain-boundary diffusion enhancement factor on the related fission gas behaviors.

Modified Elastic Solution Processes and Interchangeable Elasticity Parameters in the Problems of Radiation Creep

Modified Elastic Solution Processes and Interchangeable Elasticity Parameters in the Problems of Radiation Creep

Development of a damaged viscoplasticity model with preliminary application in Zircaloy cladding material

Structural integrity of nuclear fuel cladding is the most important factor limiting the life and the safety of nuclear fuels. However, a rigorous evaluation of the damage and fracture behavior of the cladding is hindered by the absence of a constitutive model which can reasonably describe the complex mechanical behavior of the metallic cladding material. In this paper, a damaged viscoplasticity model is proposed by combing the damaged plasticity model, i.e., the classical Gurson-Tvergaard-Needleman (GTN) model, and a general viscoplasticity model. After instantiating the general viscoplasticity model by the thermal creep, the irradiation creep and the irradiation growth models, the damaged viscoplasticity model can thus describe the damaged viscoplastic behavior of Zircaloy cladding material. In addition, the model can describe the hydrogen embrittlement by correlating the damage parameters with the hydride content. A stress updating algorithm is proposed to implement the model within the general finite element software, i.e., ABAQUS. Then the model is evaluated by parametric studies. The typical failure behavior of the Zircaloy cladding is also investigated by using the developed constitutive model.

Polycrystalline simulations of in-reactor deformation of recrystallized Zircaloy-4 tubes: Fast Fourier Transform computations and mean-field self-consistent model

Fuel cladding and structural components made of zirconium alloys, used in light and heavy water nuclear reactors, exhibit, during normal operation, significant in-reactor deformation. Fast Fourier Transform (FFT) simulations have been conducted on large grain aggregates to simulate the in-reactor behavior of recrystallized Zircaloy-4. Original constitutive equations have been proposed to account, at the microscopic scale, for thermal creep, irradiation creep and irradiation induced growth. The evolution of irradiation defects with irradiation is taken into account, especially to deduce the local growth strain. A good description of the in-reactor behavior is obtained with irradiation defects evolution consistent with Transmission Electron Microscopy observations. The FFT simulations are compared to a self-consistent model. A good agreement is obtained when the behavior is linear (irradiation creep and growth) while the nonlinear response (thermal creep) is underestimated by the self-consistent model. The FFT simulations are also compared to the lower-bound model which neglects the interactions between grains. The lower-bound model underestimates the growth strain proving the importance of using an accurate polycrystalline model to predict the growth strain from the knowledge of the irradiation defect evolution.

Core Materials for Sodium-Cooled Fast Reactors: Past to Present and Future Prospects

Abstract The Sodium-Cooled Fast Reactor (SFR) is the most promising and technologically evolved among the six nuclear reactor concepts selected by the Generation IV International Forum. Although it is compatible with a closed fuel cycle for sustainable energy production, its competitiveness depends on being able to achieve high fuel burn-up up to 200 GWd/t. This would enable efficient fuel utilization to minimize fuel cycle costs. Hence, development of optimized core materials, especially for fuel cladding tubes that are subjected to extreme conditions of intense fast-spectrum neutron irradiation, high temperatures, and mechanical and chemical fuel–cladding interactions, continues to be a priority worldwide. Resistance to void swelling, irradiation creep, and embrittlement are required to be enhanced to minimize dimensional changes and loss of ductility in core components, which limit achievable burn-up. Current SFR core materials comprise austenitic stainless steels (SS) and ferritic-martensitic (FM) steels. Cold-worked austenitic SS grades 304, 316, 316Ti, niobium-stabilized grades FV548, EI-847, EP-172, and Ti-modified 15Cr-15Ni SS, such as D9, 1.4970, 15/15Ti SS, and ChS-68, have been used for fuel cladding and ducts in SFRs built to date, and they withstand neutron irradiation damage up to 80–100 dpa. Subsequent improvements have been made by optimizing minor elements (titanium, silicon, and phosphorous) and multi-stabilization to develop advanced grades such as Si-modified 15-15Ti, Indian Fast Reactor Advanced Clad (IFAC-1), PNC-316, EK-164, and HT-UPS, which could support burn-up of 150 GWd/t. In the case of FM steels, several commercial grades are found suitable in view of their inherent void swelling resistance to 200 dpa, and are being developed for improved creep strength by oxide dispersion strengthening to realize burn-up of 200–250 GWd/t. This article presents the development of structural materials for SFR fuel pin cladding and ducts and associated enhancement of permissible fuel burn-up. Indian and international experience resulting from extensive in-pile and out-of-pile mechanical testing and fuel–cladding interactions has been covered.

A constitutive framework for steady-state creep in coarse-grained zirconium alloys under irradiation

ABSTRACT A constitutive framework for steady-state creep in polycrystalline zirconium alloys under neutron irradiation is elaborated. The framework combines a two-scale description with internal variables. The structure of the local constitutive laws within individual grains follows from a thermodynamical formulation accounting for freely migrating defects nucleated by radiation and for deformation produced by dislocation glide and climb, while the structure of the overall constitutive relations for the polycrystalline aggregate follows from homogenisation. The resulting description is thermodynamically consistent and conforms to the generalised standard material model. A special class of constitutive laws is generated using convex dissipation potentials reproducing the mean-field reaction rate theory. The relations are nonlinear and reflect an inherent coupling between irradiation creep and growth; however, they admit partial linearisation even for moderate stress levels. Some implications for mechanistic creep models of common use are discussed.

A molecular dynamics study of a cascade induced irradiation creep mechanism in pure copper

Recently, in-situ TEM straining experiments on pure copper have unraveled a high stress irradiation creep mechanism. Irradiation induced unpinning of dislocations from defects has been observed and the mean pinning lifetime has been determined. In the present study, molecular dynamics simulations are performed on pure copper to investigate the impact of collision cascades on screw dislocations pinned on Frank loops under high-applied stresses at 300 K in order to further quantitatively elucidate this mechanism. The simulations indicate two possible dislocation unpinning mechanisms. Unpinning can occur through loop destruction when the cascade is generated on the pinning points of the dislocation (type 1 unpinning). Unpinning can also be triggered by the shear stresses building up around a cascade generated in front of the dislocation in the glide plane (type 2 unpinning). Type 2 unpinning generally leads to dislocation repinning on cascade residues, so that it should only marginally contribute to irradiation creep. The mean pinning lifetime due to type 1 unpinning in the conditions of the in-situ TEM experiments is derived from a simple model previously developed for zirconium, and is found in the same orders of magnitude as in the experiments.