Irradiation Growth Research Articles

Effect of temperature on the neutron irradiation-induced cavities in tungsten

ABSTRACT To study the effect of temperature on the evolution of irradiation-induced cavities in tungsten, ITER Grade Product tungsten was characterised using transmission electron microscopy (TEM) after neutron irradiation to ∼1 dpa at 600°C, 800°C, and 1000°C. The results showcase irradiation temperature-dependent cavity formation and growth, which is majorly governed by the spatial distribution of transmutation elements in the matrix. Suppression of thermal neutrons fluence by applying stainless steel capsules resulted in a lower density of secondary phases as well as cavities than those reported in the literature for similar irradiation conditions. To further assess temperature effects on the cavity evolution and their stability, dedicated in situ TEM post-irradiation annealing (PIA) experiments were carried out. PIA of 600°C irradiated samples at 600°C for about 1 h revealed no noticeable change on the cavities size distribution. However, upon increasing annealing temperature and time, cavities became coarsened. Vacancy migration and absorption is found to be the dominant mechanism that governs cavity annealing kinetics in the present thin-foil PIA investigations. Furthermore, the cavity coarsening is dictated by the simultaneous evolution of the irradiation-induced dislocation loops, thin-foil free surfaces image forces, and the presence of the transmuted elements and their clusters in the matrix.

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.

Irradiation growth behavior and effect of hydrogen absorption of Zr-based cladding alloys for PWR

In order to understand the dimensional stability of the fuel rod during long-term use in commercial LWRs, irradiation growth testing on various Zr-based fuel cladding materials including improved Zr-based alloys was conducted in the Halden reactor in Norway under prototypic irradiation and corrosion environment. The effect of hydrogen absorbed in the cladding tube due to corrosion on the irradiation growth behavior was investigated in this paper. The comparison of specimens with or without pre-charged hydrogen (about 200 and 400 ppm) revealed that the effect of hydrogen absorption, i.e. the acceleration of irradiation growth, became significant when the hydrogen content exceeded the hydrogen solubility limit in the corresponding irradiation temperature. Based on this understanding, the growth acceleration effect was evaluated to be (0.06 ± 0.01)%/100 ppm, with the denominator defined as the amount of absorbed hydrogen involved in hydride precipitation under irradiation as a relevant parameter. Discrimination of hydrogen effect confirmed lower irradiation growth of the improved Zr-based alloys than the conventional Zry-4, which may be attributed to the effect of lower hydrogen absorption property for Nb addition specimens.

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.

Microstructural complexity and dimensional changes in heavily irradiated zirconium

Using atomistic simulations based on the creation-relaxation algorithm, we explore the evolution of microstructure in irradiated zirconium over a broad range of radiation exposure. In agreement with experimental observations, we find that at relatively low temperatures, microstructure evolves towards an asymptotic dynamic steady state forming at doses close to 1 dpa. Simulations show the spontaneous formation of $a$-type interstitial dislocation loops, gradually transforming into an $a$-type extended dislocation network and giving rise to macroscopic anisotropic dimensional changes in a textured material. A fully developed $a$-type interstitial/vacancy and $c$-type vacancy dislocation microstructure corresponds to the highest degree of irradiation growth anisotropy and vanishingly small volumetric swelling of the material.

Determination of Thermal Expansion, Defect Formation Energy, and Defect-Induced Strain of α-U Via ab Initio Molecular Dynamics

Uranium (U) is often alloyed with molybdenum (Mo) or zirconium (Zr) in order to stabilize its high-temperature body-centered cubic phase for use in nuclear reactors. However, in all metallic fuel forms, the α phase of U remains in some fraction. This phase decomposition due to temperature or compositional variance can play an outsized role on fuel performance and microstructural evolution. Relatively little is known about fundamental point defect properties in α-U at non-zero temperatures, from either computational or experimental studies. This work performs the first thorough evaluation of the α phase of U via ab initio molecular dynamics (AIMD). A number of thermophysical properties are calculated as a function of temperature, including equilibrium lattice parameters, thermal expansion, and heat capacity. These results indicate a two-region behavior, with the transition at 400 K. The thermal expansion/contraction in the a/b direction occurs rapidly from 100 up to 400 K, after which a more linear and gradual change in the lattice constant takes place. The volumetric expansion matches experiments quantitatively, but the individual lattice constant expansion only matches experiments qualitatively. Point defect formation energies and induced lattice strains are also determined as a function of temperature, providing insight on defect populations and the fundamentals of irradiation growth in α-U. Interstitials induce significantly more strain than vacancies, and the nature of that strain is highly dependent on the individual lattice directions. The direction of point defect-induced lattice strain is contrary to the irradiation growth behavior of α-U. This work shows that isolated point defects cannot be the primary driving force responsible for the significant irradiation-induced growth of α-U observed experimentally.

Endoleak Detection after Endovascular Aortic Aneurysm Repair Using Ultrasound Based on Nanoscale Bubble Contrast Agents and Their Effects on Vascular Smooth Muscle Cell Proliferation and Migration

Abdominal aortic aneurysm (AAA) is the most common vascular disease that causes disability and death. Its morbidity is relatively subtle, and the mortality rate is high. Clinically, endovascular aortic aneurysm repair (EVAR) has gradually become the primary treatment of AAA due to its unique advantages such as low trauma and low incidence of short-term complications. However, the outcome of EVAR is greatly compromised by the possible occurrence of endoleaks. Contrast-enhanced ultrasound (CEUS) is a promising alternative technique to detect endoleaks following EVAR due to lack of exposure to ionizing radiation. Traditional ultrasound contrast agents with an overlarge size (microscale) leading to reluctant accumulation in target organs and instability trigger the requirement of nanoscale contrast agents that enter tumor tissues through the enhanced permeability and retention effect. In this study, we used ultrasound based on nanoscale bubble contrast agents to evaluate endoleak detection after endovascular aortic aneurysm repair and analyzed the effects of nanoscale bubble contrast agents on vascular smooth muscle cell (VSMC) proliferation and migration. Among 52 AAA patients a month following EVAR, there were 16 cases of endoleaks after EVAR detected by nanobubble contrast-enhanced ultrasound, including 6 cases of type I endoleak (1 case of type Ia endoleak and 5 cases of type Ib endoleak), 7 cases of type II endoleak, and 3 cases of type III endoleak; there were 12 cases of endoleak after EVAR detected by computed tomography angiography (CTA), including 6 cases of type I endoleak (1 case of type Ia and 5 cases of type Ib), 5 cases of type II endoleak, and 1 case of type III endoleak. Six months after EVAR, 3 cases of type III endoleak were detected by both nanobubble contrast-enhanced ultrasound and CTA. Vascular smooth muscle cells (VSMCs) used for in vitro experiments were subjected to ultrasound irradiation and platelet-derived growth factor (PDGF) treatment with or without the addition of nanobubble contrast agents. After high-intensity and long-term irradiation (0.75 W/cm2 and 1 W/cm2 irradiation for 120 s and 150 s, respectively) by ultrasound with or without the addition of nanobubble contrast agents, PDGF-induced VSMC migration was inhibited ( P < 0.01 ). Low-intensity and short-term ultrasound irradiation did not differ PDGF-induced VSMC migration ( P > 0.05 ), but 0.5 W/cm2 and 90 s ultrasound irradiation could significantly inhibit PDGF-induced VSMC migration without the addition of nanobubble contrast agents ( P < 0.05 ). When VSMCs were irradiated at 1, 0.75, 0.5, and 0.35 W/cm2 for 30 s to 150 s, ultrasound irradiation with or without the addition of nanobubble contrast agents remarkably reduced PDGF-induced VSMC proliferation, as evidenced by reduced OD values ( P < 0.05 ). In conclusion, ultrasound based on nanoscale bubble contrast agents is an effective alternative detection method for the occurrence of AAA patients who are not suitable for CTA.

Modelling precipitate amorphization in Zircaloys

The second phase particles (SPPs) that occur in Zircaloys can undergo amorphization and loss of iron in-reactor, which can have a profound effect on corrosion resistance and irradiation growth. In this paper, a classical model is presented that predicts the interaction between these processes, based on a previously proposed physical mechanism that considers iron release and transport from the SPP as the controlling factor. It is assumed that iron is released into a mobile state by a homogeneous and continuous amorphization and recrystallization process. As iron is released, it diffuses out of the SPP, causing the relative stability of the amorphous phase to increase, until a persistent amorphous layer forms. The model predicts the growth rate of this layer and the rate at which iron is lost from the SPP. It has been used to explore the effect of flux, temperature, and initial SPP iron concentration on the amorphous layer growth and iron loss kinetics. It is predicted that whilst iron loss is correlated with the thickness of the amorphous layer, the relationship is not a simple scaling, and depends on conditions. The model is also able to explain the observed occurrence of an amorphous layer only in an intermediate temperature range as a natural consequence of the competition between the same fundamental mechanisms.

APT characterization and modeling of irradiation-induced Nb-rich nanoclustering in Zr-1.0%Nb alloys

Recent development of Zr-based alloys containing Nb solutes have demonstrated promising improvements in corrosion resistance and reduced irradiation growth, particularly when irradiation-induced Nb-rich nanoclusters are present and inhibit 〈c〉 loop formation. In this study, a Zr-1.0%Nb alloy is irradiated with either Kr2+ ions or neutrons to 5 dpa at 310 °C. Using atom probe tomography, radiation-induced Nb-rich nanoclusters and a corresponding reduction of Nb matrix composition from 0.40 at% to 0.26 at% and 0.34 at% is characterized following each irradiation, respectively, suggesting that matrix solutes coalesce to form nanoclusters. Irradiation with Kr2+ ions induces nanoclusters that are larger and denser than those following neutron irradiation. Utilization of a simple rate-theory nanocluster evolution model demonstrates high sensitivity to the solute migration, vacancy formation, and vacancy migration energies directly influencing radiation-enhanced mobility of solutes, enabling exploration of irradiation temperature effects and potential differences in nanocluster evolution for different solutes or alloy systems.

Coupled Thermomechanical Responses of Zirconium Alloy System Claddings under Neutron Irradiation

Zirconium (Zr) alloy is a promising fuel cladding material used widely in nuclear reactors. Usually, it is in service for a long time under the effects of neutron radiation with high temperature and high pressure, which results in thermomechanical coupling behavior during the service process. Focusing on the UO2/Zr fuel elements, the macroscopic thermomechanical coupling responses of pure Zr, Zr-Sn, and Zr-Nb binary system alloys, as well as Zr-Sn-Nb ternary system alloy as cladding materials, were studied under neutron irradiation. As a heat source, the thermal conductivity and thermal expansion coefficient models of the UO2 core were established, and an irradiation growth model of a pure Zr and Zr alloy multisystem was built. Based on the user material subroutine (UMAT) with ABAQUS, the current theoretical model was implemented into the finite element framework, and the consequent thermomechanical coupling behavior under irradiation was calculated. The distribution of temperature, the stress field of the fuel cladding, and their evolution over time were analyzed. It was found that the stress and displacement of the cladding were sensitive to alloying elements due to irradiated growth.

A new irradiation growth model for Zr-based components of nuclear reactors for the DIONISIO code

When zirconium-based materials are subjected to irradiation, dimensional changes occur due to irradiation creep and growth. Predicting the behavior of these phenomena is necessary to determine the maximum time of operation required to avoid possible mechanical problems, such as buckling of sheaths, cracks, sheath-cooling channel contact, etc. In this work, a model to evaluate the irradiation growth based on the evolution of microstructural defects such as dislocations and point defects is presented. Preliminary results obtained are in good agreement with experimental data. In addition, the model was incorporated as a new subroutine within the nuclear fuel code DIONISIO and some comparisons with experimental data and previous models in DIONISIO are presented.

Two-dimensional vacancy platelets as precursors for basal dislocation loops in hexagonal zirconium

Zirconium alloys are widely used structural materials of choice in the nuclear industry due to their exceptional radiation and corrosion resistance. However long-time exposure to irradiation eventually results in undesirable shape changes, irradiation growth, that limit the service life of the component. Crystal defects called <c> loops, routinely seen no smaller than 13 nm in diameter, are the source of the problem. How they form remains a matter of debate. Here, using transmission electron microscopy, we reveal the existence of a novel defect, nanoscale triangle-shaped vacancy plates. Energy considerations suggest that the collapse of the atomically thick triangle-shaped vacancy platelets can directly produce <c> dislocation loops. This mechanism agrees with experiment and implies a characteristic incubation period for the formation of <c> dislocation loops in zirconium alloys.

Stereotactic Radiation Enhances Efficacy of CAR-T Cells in a Murine Glioma Model

T cell tumor infiltration is an important driver of antitumor immune response and tumor cell killing in immune checkpoint inhibitor (ICI) therapy. Many modalities have been investigated to increase T cell tumor infiltration and thereby enhance antitumor immune response including cytotoxic therapies such as radiation. Radiotherapy promotes local inflammation, increases tumor antigen release, and can enhance T cell priming leading to a more robust T cell immune response. Many studies have investigated the combination of radiotherapy and ICI therapy, but limited work has been done examining the combination of targeted CAR-T cell therapy with the T cell immune enhancing effects of radiotherapy. To explore this question, we investigated the anti-tumor efficacy of intra-cranial IL13Rα2 directed CAR-T in combination with tumor irradiation in an immune-competent murine glioma model. Using an immunocompetent murine glioma model, we first determined the sub-therapeutic dose of radiation for our model by using increasing doses of single fraction irradiation (5 Gy, 10 Gy, and 20 Gy). We performed gene-expression panel analysis on brain tumor tissue at each dose to evaluate for changes in tissue expression of chemo-attractant molecules following irradiation. After identifying an appropriate subtherapeutic radiation dose, we tested the combination efficacy of focal radiation and I.C.T delivered IL13Rα2 CAR-T in the glioma model by injecting I.C.T IL13Rα2 CAR T cells 48 hours after 10 Gy tumor irradiation and monitoring tumor growth and response. We then assessed immunologic memory after therapeutic cure by focal radiation, I.C.T. delivered IL13Rα2 CAR-T cell therapy, and combination therapy by re-challenging the cured mice with intracranially implanted IL13Rα2 (+) tumors. In an immunocompetent murine glioma model single fraction doses of 5 Gy and 10 Gy were found to be subtherapeutic. mRNA tissue analysis following irradiation demonstrated statistically significant upregulation of mRNA encoding chemoattractant molecules such as Cxcl2. Mice treated with focal radiation and I.C.T. delivered IL13Rα2 CAR-T exhibited improved survival when compared to mice treated with mock CAR-T cell injection followed by irradiation or with CAR-T cell therapy alone. Mice previously cured with 10 Gy and IL13Rα2 CAR-T treatment demonstrated immunologic memory and rejected IL13Rα2 (+) tumor re-challenge in 5 of 5 mice as compared to naïve mice. Irradiation upregulates mRNA expression of chemoattractant molecules in an immunocompetent murine glioma model. Combination treatment with focal radiation and I.C.T. IL13Rα2 CAR T cell therapy shows improved survival compared to CAR T cell therapy or radiation alone. Combination therapy also leads to enhanced immune memory and improved tumor rejection with tumor re-challenge.

A surrogate model for the 3D prediction of in-service deformation in CANDU® fuel channels

In CANDU reactors, a contact between the Pressure Tube (PT) and the Calandria Tube (CT) could cause the development of hydride blisters that may lead to delayed hydride cracking of PTs. To maintain PT integrity, contact between these two tubes must be avoided. Under normal operating conditions (NOCs), the PT-CT contact can occur as a result of in-reactor deformation due to irradiation induced creep, irradiation growth and thermal creep of the fuel channel (FC) assembly. Numerical models have been developed to predict the time of PT-CT contact in a channel, which is required to ensure the operability of the channel and the plan for maintenance. Since the prediction of time to contact is influenced by various uncertainties, such as change in, (i) the dimensions of the FC, and (ii) the material properties and boundary conditions of the FC, probabilistic simulation-based methods have been developed to assess the PT-CT contact risk and establish adherence with provisions of the Canadian Standards Association (CSA) Standard N285.8. Numerical simulations based on a 3-dimensional finite element model (3D FEM) are essential in predicting the dimensional changes occurring during in-reactor deformation of single FCs because, unlike 1-dimensional FEM, 3D analyses correctly take into account the triaxial loading conditions in the calculation of the complete strain tensor. However, 3D FEM analyses are not practical for probabilistic assessments of an entire reactor core with 380 or 480 FCs. This paper (i) outlines the steps that are required to develop a 3D model of a FC using the ABAQUS commercial code along with user subroutines that incorporate the constitutive equations describing the in-reactor deformation of FC materials, and (ii) proposes a new surrogate model of a much simpler analytical form, which can replace the 3D FEM for assessing PT-CT contact in carrying out whole core probabilistic assessments. The proposed surrogate model not only simplifies the FE analysis and provides a more rational basis for probabilistic assessments, but most importantly allows the inclusion of diametral expansion and wall thinning in the prediction of PT-CT gap. This simple function representation gives a cost-effective and computationally efficient approach to probabilistic analysis of FCs.

Assessment of WWER-1000 Core Baffle Form Alteration during Operation

The paper illustrates the results of the computer assessment of the form alteration in WWER-1000 core baffle obtained via the solution to the coupled thermoelastoplastic task considering the strains of irradiation growth and creep. In the modeling of the contact conditions, the temperature redistribution is considered due to the incompliance of the coolant flow in the contact zone between the core baffle and in-vessel core barrel with the design conditions. The modern approaches to the modeling of strains of the irradiation growth and irradiation creep in austenite steels are used in the space-limited environment under neutron exposure and elevated temperature. The finite element analysis involves the mixed scheme of the finite element method, which allows determination of the stress-strain state with high accuracy. The calculations are performed in the two-dimensional statement for the cross-section of the core baffle with the maximum damaging dose and irradiation temperature under the condition of the generalized plane strain. The results of the calculations are presented for full-scale reactor operation and scheduled shutdown to recharge the fuel cluster at the end of core life. The data on the distribution and value of the gap between the core baffle and barrel, as well as the spacer grids of the edge fuel assemblies and reactor core baffle edges, have been obtained from the median values of the dose dependence on swelling at different temperatures in Kh18N10T austenite steel.

Cluster dynamics modeling of irradiation growth in single crystal Zr

Irradiation growth of single-crystal Zr is modeled with a reduced set of cluster dynamics equations. Nucleation and growth of basal and prismatic dislocation loops are both accounted for in the model. Equations are developed for the time evolution of single point defects, a limited number of point defect clusters, interstitial and vacancy loop nucleation rates, as well as for the growth of vacancy loops on basal planes and interstitial loops on prismatic planes. Reduction of the usual infinite hierarchy of cluster dynamics equations to the simple set studied here is justified on the physical basis of the stability of small loops once nucleated. This simplified cluster dynamics model with a small number of adjustable parameters avoids the complexity of explicit representation of higher order point defect clusters. The model shows consistency with experimental observations of the following aspects: (1) the growth rates of Zr crystals along the a- and c-axes; (2) the onset dose for breakaway irradiation growth; (3) the saturation dislocation loop densities of vacancy and interstitial loops; (4) the effects of cold work and temperature on irradiation growth.

The effect of local chemical environment on the energetics of stacking faults and vacancy platelets in [formula omitted]-zirconium

The onset of breakaway irradiation growth in Zr and its alloys has been correlated with the nucleation and growth of faulted vacancy c-loops on basal planes. One theory for c-loop stabilization at high fluence is the reduction in stacking fault energy by solute or impurity segregation. This hypothesis is investigated through density functional theory calculations of the binding energy of Fe, Sn, Nb, Cr, and Ni to basal extrinsic, basal intrinsic or prismatic stacking faults in α-Zr. By means of a thermodynamic model, the ab initio binding energies have been correlated to segregation profiles and modified SFEs. Sn and Fe are shown to be tightly bound to basal stacking faults with only a weak interaction towards prismatic stacking faults. The corresponding segregation to faulted basal c-loops significantly reduces their formation energies and has the potential to stabilize these defects in the matrix. Based on the DFT calculations presented here, Nb, Cr, and Ni do not have a considerable effect on faulted c-loop formation energies. The binding energies of Fe, Sn, Nb, Cr, Ni, and H to a 19-vacancy platelet were additionally calculated to confirm the behavior of these elements with a nucleating c-loop defect. Fe and Cr in particular show strong binding to the periphery of this high-energy structure and correspondingly reduce the nucleation barrier for a vacancy c-loops.

Knockdown of annexin VII enhances nasopharyngeal carcinoma cell radiosensitivity in vivo and in vitro.

Radioresistance leads to treatment failure in patients with nasopharyngeal carcinoma (NPC). Thus, enhancing the radiosensitivity of NPC cells would likely increase the effectiveness of radiotherapy. Annexin VII (Annexin A7, ANXA7) might be a tumor promoter in NPC but its functions in radiosensitivity remain unclear. NPC celllinesCNE2-shANXA7 and CNE2-pLKO.1 were generated and CNE2-shANXA7 nudemicexenograft tumormodels were established. The main effects and molecularmechanisms of ANXA7 knockdown in NPC radiosensitivity were studied in vitro and in vivo by analyzing cell viability, clonogenicity, apoptosis, cell cycle distribution, tumor radioresponse and immunohistochemistry assay. ANXA7 knockdown revealed potentially enhanced NPC cell radiosensitivity via apoptosis and increased the cell numberat the G2/M phase. In the xenograft model, NPC cells with ANXA7 knockdown were dramatically sensitive to irradiation and tumor growth was significantly suppressed. Compared to CNE2-pLKO.1 xenografts, CNE2-shANXA7 showed more γ-H2AX foci and less phospho-DNA PKcs. ANXA7 knockdown increased the radiosensitivity of NPC by enhancing apoptosis, modulating the cell cycle distribution into more radiosensitive phases, promoting DNA damage, and inhibiting repair. We showed that decreased ANXA7 levels enhanced radiosensitivity and provided insights into the therapeutic targets for NPC radiotherapy.

Nuclear Fuel Assembly Deformation, Reduced Mechanical Model Dedicated to FSI Simulation

In the pressure vessel of a PWR the core is constituted of a number of fuel assemblies. Under operation, they all suffer from various phenomena, among them, initial trapped efforts, thermal expansion, irradiation growth, creeping, tightening relaxation and rod slipping. Each fuel assembly may develop contacts with its neighbors and strongly interacts with the fluid. It then undergoes vertical as well as lateral fluid forces, which are partly coupled with its instantaneous shape. As a consequence the fuel assembly tends to get distorted during and still after operation. The first concern is economic: at the end of a cycle, when changing the assembly position in the core, the contact interactions with its neighbors due to its distorted shape sometimes delay the handling. The second concern is that the shutdown rods have to slide into the deformed shape of the guide tube and might be slowed due to excessive friction forces. In order to precisely assess these situations it is necessary to simulate numerically a complete core, taking into account all of the previously mentioned phenomena. For that purpose, we propose a light but relevant reduced mechanical model of a fuel assembly, partially based on a POD analysis performed on a detailed fuel assembly model. We then show that a rather simple model, still based on classical finite elements can be used and simplifies the integration of the local nonlinear constitutive equations. We then compare both of the detailed and reduced model on typical situations and show excellent agreement. The CPU time ratio between them, for a full nonlinear simulation of a single fuel assembly under operation, is over 1000, which is competitive against separation of variables based reduced order methods such as POD, PGD and even APHR. So, yet still with a simplified FSI approach, the drastic memory and CPU reductions obtained now enable us to complete a 3D core simulation representing at least 4 years under operation (4 cycles), with as much as 241 individual fuel assemblies, in a reasonable time of a few hours.