Fuel Rod Temperature Research Articles

Understanding post-irradiation creep behavior of M5Framatome zirconium alloy

After in reactor use, the nuclear fuel assembly of pressurized water reactors is stored in water and is eventually transported, in a dry cask, to a long-term storage facility or to a reprocessing plant. During transportation in dry environment the fuel rod temperature and internal pressure both increase. This leads to simultaneous creep deformation and annealing of the cladding tubes made of M5Framatome11M5 and M5Framatome are trademarks or registered trademarks of Framatome or its affiliates, in the USA or other countries. zirconium alloy. This study shows that post-irradiation creep, or heat treatment, leads to the recovery of the radiation induced hardening and also to the recovery of the uniform elongation because of <a>-loop annealing. On the other hand, it is shown that cladding tubes made of M5Framatome alloy exhibit low post-irradiation creep rates, especially at low applied stress. This strain rate is significantly lower than for the unirradiated material and similar to the as-irradiated material, despite the significant <a> loop annealing. The low creep rates are mainly attributed to the Nb-rich nano-precipitates that act as dispersed strengthening particles. These nano-precipitates appear under irradiation and are stable at temperatures up to 450 °C explaining their strengthening effect even after significant annealing.

Performance analysis of irradiated thermo-mechanical coupled petal-shaped fuel rod under different power and heat transfer conditions

As a new type of fuel rod, petal-shaped fuel rod (PSFR) has the advantages of lowering the core temperature and increasing the power density, and thus has a broad application prospect. Based on the thermal elastic–plastic theory, this paper establishes an analytical model for irradiated thermal–mechanical coupling performance of PSFR by user subroutine. Through numerical simulations of the full-cycle operation of PSFR under different power and heat transfer conditions, the performance parameters of fuel rods, such as fuel rod temperature and stress, are obtained, and their spatial distribution and the variation pattern with the operation time are analyzed. The results show that increasing line power leads to a simultaneous increase in fuel rod temperature, stress, maximum deformation, and creep. However, an appropriate reduction in heat transfer coefficient decreases the cladding stresses. The above research results will provide an essential theoretical basis for determining new PSFR’s deformation and operation risk.

Investigations of Multiphysics Models on a Megawatt-Level Heat Pipe Nuclear Reactor Based on High-Fidelity Approaches

This work describes the research of high-fidelity multiphysics models for the MegaPower nuclear reactor, a megawatt-level heat pipe reactor. Combining the Monte Carlo neutronics model, the heat pipe analysis model, the fuel analysis model, and the thermoelasticity model produces the Multi-Physics Coupling code for Heat pipe nuclear reactors (MPCH) code platform. Using the heat pipe analysis model, a database of heat pipes is generated to save computing costs. Comparison is made among four calculating modes with differing degrees of coupling. It was discovered that the thermal expansion effect reduces core reactivity by 537 ± 11 pcm and the temperature feedback coefficient by 61%. With the incorporation of the heat pipe module, a temperature difference arises between the wall of heat pipes, which can reach a maximum value of 80 K at steady state. Simultaneously, the global fuel rod temperature difference increases from 34 K (under the assumption of uniform heat pipe wall temperature) to 93 K, and the monolith temperature variance increases from 34 to 108 K. At the periphery of the monolith, the increased temperature variation causes a monolithic stress of 188.6 MPa. To further investigate the safety of the reactor, three-heat-pipe-failure scenarios are evaluated. The heat pipe analysis model reveals that a single heat pipe failure results in a monolith peak temperature of 1046 K, giving a maximum monolith stress of 237 MPa. The maximum monolith stresses and temperatures for the two-heat-pipe-failure scenario and the three-heat pipe-failure scenario are 330 MPa/1128 K and 471 MPa/1233 K, respectively. In steady-state operation, the stresses exceed the yield tensile strength (131MPa) whereas those generated by the failure of three heat pipes exceed the ultimate tensile strength (345 MPa) in high temperature. These results illustrate the necessity of including coupled multiphysics models into the design and safety evaluation of innovative nuclear reactors.

Modeling and program development of supercritical CO2 sub-channel based on Modelica language

The application of supercritical carbon dioxide (S-CO2) in small nuclear reactors is a new research direction with promising prospects. Most of the research on S-CO2 cooled reactors focuses on the design and optimization of core parameters, but few on detailed thermal and hydraulic analysis of the reactor interior using the sub-channel method. To supplement the lack of research in S-CO2 small nuclear reactors, this paper developed an S-CO2 cooled reactor two-dimensional sub-channel analysis model based on Modelica language. To verify the correctness of the model and program, the calculated results of the sub-channel program are compared with the results of KAIST MMR in Korea, the power deviation is less than 4%, the coolant temperature deviation is less than 7%, and the fuel rod temperature deviation is less than 10%. Based on the developed sub-channel model, the transient analysis is carried out, including reactor power increase, reactivity insertion, and coolant flow loss accidents, and then the safety performance of the reactor is evaluated. The transient response of the reactor under accident conditions is obtained, and the self-stability and self-adjustment characteristics are proved. This paper relies on theoretical modeling and program development methods, studying the sub-channel thermal-hydraulic characteristics of steady-state and transient conditions, which provides theoretical reference and program support for the study of the S-CO2 small nuclear reactor.

CorTAF: A nuclear reactor core three-dimensional thermal-hydraulic characteristics analysis code based on OpenFOAM

In this paper, a thermal hydraulic analysis method based on open source CFD platform OpenFOAM was proposed. The models of coolant flow and heat transfer, fuel rod heat conduction and coupled heat transfer were established according to the bundle structure characteristics of PWR core, and then a nuclear reactor three-dimensional thermal–hydraulic characteristics analysis code CorTAF was developed based on finite volume method with the spatialresolution of subchannel level. The international benchmarks, including GE3 × 3, Weiss and PNL2 × 6 fuel assembly flow and heat transfer experiments, were selected to carry out the validation. The calculation results were basically consistent with the experimental data, illustrating that CorTAF is suitable for predicting the flow and heat transfer characteristics of coolant in the rod bundle fuel assembly. The thermal–hydraulic characteristics of fuel assembly and PWR core under full power operation and blockage accident conditions were simulated using CorTAF. The steady-state and transient spatial distribution of physical quantities in subchannel-scale including coolant and fuel rod temperature were obtained, and the influence of blockage condition on flow and heat transfer characteristics was analyzed. This work has reference significance for the development of core thermal hydraulic analysis tools for PWR.

Application of machine learning algorithm in the prediction of fuel rod temperature performance

In order to predict the fuel rod temperature performance effectively, we establish a series of machine learning models based on <italic>k</italic>-nearest neighbor, decision tree and AdaBoost algorithms. The input parameters and calculation results of fuel rod performance analysis software JASMINE and data feature engineering are used as the training and test data for the machine learning models. The models are trained by the training data set which includes the characteristic parameters such as pellet and cladding type, axial height, local power, cladding corrosion thickness and core inlet temperature. After training, the models use the test data to predict the cladding outside surface temperature and pellet center temperature. The prediction results show that the model based on AdaBoost algorithm has the best prediction performances, and the mean square errors of cladding outside surface temperature and pellet center temperature are 0.605 ℃ and 8.347 ℃, respectively, and the average absolute errors are 0.273 ℃ and 3.814 ℃, respectively. Comparing the predicted values with the target values, the maximum deviation of Adaboost algorithm for the cladding outside surface temperature is 3 ℃, and the most of the prediction deviation of the pellet center temperature is less than 10 ℃, indicating that the model based on AdaBoost algorithm has the high prediction accuracy for the temperature performance of fuel rods.

Preliminary development of a multi-physics coupled fuel performance code for annular fuel analysis under normal conditions

Benefitted from its dual-cooled structure, annular fuel may enhance fuel performance against conventional solid fuel. To aid the potential deployment of annular fuel, fuel performance analysis can be beneficial. In this study, leveraging the open-source Multiphysics Object-Oriented Simulation Environment (MOOSE), a finite element method (FEM) based fuel performance code, have been preliminarily developed for annular fuel, where governing equations for both fuel and coolant are solved in a fully coupled manner. Afterward, assessments of the developed codes were conducted through selected cases to improve the confidence of code predictions. Then, a comprehensive analysis of annular fuel and its fuel performance comparison against solid fuel were completed. Meanwhile, the uncertainty of fuel relocation on fuel performance was investigated. The results suggest that the (1) predictions by the developed code are reasonable, (2) fuel rod temperature & fission gas release can be notably reduced even under elevated rod power (∼1.2x solid fuel power) when annular fuel is employed, (3) the uncertainty of fuel relocation plays a vital role for fuel heat split between internal and external coolant channels which should be considered to evaluate its thermal–hydraulic performance.

Improved Radiation Heat Transfer Model in RELAP5 for Compact Fuel Rod Bundles by the Absorption Factor Modification

Simulations of radiation heat transfer in fuel rod bundles are necessary for the thermal hydraulic design and safety analysis of open lattice gas-cooled reactors, which always operate at high temperatures. To save the computational costs, existing radiation models in system codes such as RELAP5 commonly assume each fuel rod to own the uniform radiosity over the rod surface. Previous research studies have indicated that the uniform radiosity assumption could overestimate the heat transfer flux and under-predict the maximum fuel rod temperature, and the anisotropic correction was tried by dealing with non-uniform reflected radiation. To better model the non-uniform radiosity effect, the Gehart’s method based on the non-uniform absorbed radiation is introduced in this study. By dividing the surface of each rod into six segments, the one-sixth rod view factors are derived in specific rod and near wall sections to generate the segment-to-segment absorption factors. By summarizing those segment-to-segment absorption factors, the rod-to-rod and rod-to-wall absorption factors are modified and implemented into RELAP5 to improve the radiation heat transfer model. The two-dimension radiation heat transfer problem in the nuclear fuel rod bundle is simulated in FLUENT as the benchmark and in RELAP5 for comparison. Fuel rod bundles in hexagonal arrays were investigated with various surface emissivity and pitch-to-diameter ratios (p/d). The simulations indicated that the method of rod segment division and absorption factor modification could reflect the non-uniform radiosity, and the results were related to the values of p/d and surface emissivity. The modified radiation heat transfer model in RELAP5 validated that the deviations of the maximum temperature were reduced from around 20% to 1%,3%,8% for p/d = 1.1, 1.2, and 1.3, respectively. Rod bundles with larger p/d required more radiative rods in the analyses of absorption factor modifications. The present radiation heat transfer model should be studied and tested in three-dimension cases to further prove that it is appropriate for the nuclear rod bundles.

Analysis of fuel rod performance per cycle: Temperature field, FGP release, swelling

Models for calculating gas release, pressure of gases within fuel rod cladding, fuel swelling and thermal conductivity, and fuel-cladding gap conductance are briefly described. The thermophysical condition of fuels in a reactor core is one of the main factors that determine their serviceability. The stress-strained condition of fuel claddings under design operating conditions is closely related to fuel rod temperature, swelling, gas release from fuel pellets and the mode in which they change during the cycle and transients. Aside from this, those parameters are an independent goal of studies since their ultimate values are governed by the system of design criteria. In the group of thermal physics criteria, consideration is given to maximal temperatures of fuel, FGP releases, and pressure of gases within a fuel cladding. It is shown that the micro-processes of fission gas incipience and migration in fuel under irradiation are closely related to essentially all spectrum of macro-characteristics of fuel, namely, temperature field and heat rating, fuel burn-up, pressure and composition of gas within a fuel rod, characteristics of the stress-strained condition of fuel column and cladding. The adequate computer analysis of the dynamics of FG behaviour and affected structure changes in a wide range of irradiation conditions is most important for prediction of fuel rod serviceability and, on the other hand, it may serve to corroborate the adequacy of prediction of fuel rod macro-condition characteristics using the full-scale code.

Assessment of radiation heat transfer influence on parameters of temperature fields of various design fuel rods

VVER-1000 reactors use cylindrical smooth-core fuel rods. Previously, a model to determe the fuel rod temperature field in a two-dimensional problem statement has been developed and verified. However, modelling assumptions do not consider the influence of variable thermophysical properties, radiation heat transfer, and the opening in the fuel rod on the final parameters of the temperature fields. The impact assessment is an urgent task to improve the economic efficiency of the fuel cycle and the capacity of power units. To develop models and study the features of energy release in nuclear reactors, a numerical package of thermophysical modeling COMSOL Multiphysics software is used. The simulation of temperature fields is performed based on the heat equation with an internal heat source, under the boundary conditions of the second kind at the ends of the fuel rod and the boundary conditions of the third kind on the side surface of the rod. Аn axisymmetric model in two-dimensional problem statement and a three-dimensional model of the fuel rod are developed. The temperature distribution fields are determined by the finite element method. The results of calculations of various design fuel rods are presented. The results have showen that the radiation heat transfer significantly affects the maximum fuel temperature (UO2). The impact degree of variability of thermophysical properties and radiation heat transfer is determined. It was found that the temperature characteristics under different specified conditions have a difference in the range of 15,5–282,0 K (0,8–14,4 %). The developed models are reliable and confirmed by the previously verified model, the characteristics of the fuel assembly used on the VVER-1000 units. The results presented can be used for mathematical modeling of heat transfer processes, both during the modernization of the equipment in operation, and during the development, design, and operation, which will increase the efficiency of electric energy generation at the power unit of a nuclear power plant.

Prediction of thermophysical characteristics of fuel rods based on calculations

The paper analyzes operating conditions, thermophysical characteristics were calculated as applied to WWER-1000 fuel rods in a four-year cycle for unified core. The paper gives a summary of models for calculating gas release, pressure of gases within fuel rod cladding, fuel swelling and thermal conductivity, fuel-cladding gap conductance. The thermophysical condition of fuels in a reactor core is one of the main factors that determine their serviceability. The stress-strained condition of fuel claddings under design operating conditions is closely related to fuel rod temperature, swelling, gas release from fuel pellets and the mode in which they change during the cycle and transients. Aside from this, those parameters are an independent goal of studies since their ultimate values are governed by the system of design criteria.

PIV measurements of the cross flow induced by a wrapped wire spacer

Wrapped wire spacers have been incorporated as the spacing device in advanced nuclear reactors such as the sodium-cooled fast reactor and supercritical water-cooled reactor. The cross flow induced by the wrapped wire affects the subchannel mixing and fuel rod temperature, which is a concerned issue in the thermal-hydraulic design of the fuel assembly. In the present paper, an experiment has been conducted to study the cross flow of subcritical water in a square channel containing a single tube which was helically wrapped by a wire using particle image velocimetry. The global velocity distribution and local flow features were obtained. It was found that the cross-sectional small vortexes are generated by the fluid-wall interaction as well as the wrapped wire. Local measurements using higher spatial resolution could capture more detailed flow characteristics compared to the global measurements. The magnitude of the secondary flow weakened along the cross-flow direction due to the blockage of the wire. The cross flow became stronger with the decrease of wire pitch, but this effect was more visible when the wire pitch is shorter than a certain limit. In addition, it was also observed that the cross-sectional velocity magnitude and special turbulent kinetic energy increase gradually with increasing Reynolds number. The spatially high-resolution experimental data combined with the simple geometry and clear boundary conditions can be used for the validation of CFD codes and models in predicting the wire-induced cross flow.

Effects of newly developed entrainment model for a thermal hydraulic system code (SPACE) based on breakup of large droplets in reflood

The accurate prediction of the fuel rod temperature from a peak cladding temperature to quenching is required for a thermal–hydraulic system code in the reflood phase within a pressurized water reactor. Although conventional codes have well predicted the wall temperature trends, inaccuracies with the hydraulic parameter such as the steam flow rate still remains. To predict reasonable results that reflect an actual phenomenon, we implemented the breakup of a large droplet generated in an inverted slug flow. We suggested a new method to estimate the entrainment for an Eulerian system code (SPACE). The new method was successfully verified and compared to the analytic results. After the modification, we obtained improvements on the steam flow rate, the wall temperature, PCT times, and quenching times for FLECHT-SEASET. For RBHT, it showed mixed results with a faster decrease of the wall temperature, which implies the possibility of a transition for the reflood phenomena.

FACILITY FOR HYDROGENATION AND THERMAL TESTING OF INTERNALLY PRESSURIZED DUMMY FUEL RODS

A facility allowing to conduct experiments on dummy fuel rods up to 250 mm long, at temperatures up to 700 °C is presented. The designed facility is unique in that the conditions for the tests to be conducted on it most closely resemble (except for irradiation) those of fuel rods operation, loading and storage in SFDSF. All test parameters are programmed and regulated by special sensors, which brings the experiment on fuel rod temperature effect and pressure under the cladding as close as possible to the conditions of fuel rod operation in reactor and further storage in SFDSF. Hydride Reorientation Test (HRT) was conducted on dummy fuel rod sunder internal pressure of 3...5 MPa (at room temperature) and with hydrogen concentration of 50...300 ppm in the modes that simulate SNF handling with limiting heating to 410 °С and accidents with seven 410↔300 and 410↔180 °С thermal cycles. It has been demonstrated that the effectiveness of the influence of the test conditions in the specified modes on hydride reorientation increases with increasing hydrogen concentration and tangential stresses in the dummy fuel rod claddings. It has been shown that the test samples design, control and measurement devices, as well as the parameters estimated during the test and further investigations fully meet the test requirements.

Optimal design of a Small Modular Reactor core with dual cooled annular fuel based on the neutronics and natural circulation parameters

Small Modular Reactors (SMRs) are the creative design of nuclear reactors that have received considerable attention in recent years. Since there is insufficient operational experience in SMRs, extensive research on these types of reactors may be needed to improve the current level of these systems performance. Also, one of the approaches which can help the enhancement of a reactor power is changing its fuel geometry. For this purpose, as well as decreasing the maximum fuel temperature in the reactors, the technology of annular fuels with the ability of internal and external cooling shows its importance. The nuclear SMRs with annular fuel which is cooled internally and externally has the potential to increase high power density while maintaining a sufficient safety margin. The benefit of the dual cooled annular fuel design is that heat can be transferred to the coolant at both the outer and inner channels. Hence, in this study, for the first time, dual cooled annular fuel in the Small Modular Reactors is considered and studied thoroughly. In this paper, neutronics and natural circulation parameters are calculated for an SMR with Annular Fuel which is cooled internally and externally, with changes of the internal radius. By analyzing the fuel internal radius changes in a specific range via the neutronics and computational fluid dynamics (CFD) simulation codes, the effects of power peaking factor, effective multiplication coefficient and natural circulation parameters are investigated. For the purpose of data fitting, an artificial neural network is trained using the observed data. The input consists of different internal radiuses which yields to output consisting effective multiplication factor, power peaking factor and natural circulation parameters. The Optimal geometry of annular fuel is determined using the neural network by implementing the genetic algorithms based on these neutronics and natural circulation parameters. Finally, using a simulated optimal geometry in the hot channel by neutronics and CFD simulations, thermal–hydraulic and neutronics calculations are accomplished, and these parameters are compared with the conventional SMR which uses solid (conventional) fuel. One of the main results of the evaluation is that optimal internally and externally cooled fuel in the presented SMR has a satisfactory margin for the departure from nucleate boiling (DNB) and maximum fuel rod temperature relative to the solid fuel which is used in the conventional SMRs. Indeed, fuel center temperature of conventional SMR is much more than that of presented optimal SMR. Also, presented optimal geometry increases the core effective multiplication coefficient which leads to increase in the excess reactivity. Also, reduces total power peaking factor and improves natural circulation parameters compared to the conventional SMR.

Integrity control of an RBMK-1500 fuel rod locally oxidized under a bounding reactivity-initiated accident

In 2007, the license for the second reactor unit of the Ignalina nuclear power plant was renewed considering the safety-related modifications introduced in this reactor. The Safety Analysis Report for this reactor unit was prepared with more strict criteria. The bounding reactivity-initiated accident (RIA) performed by the Lithuanian Energy Institute could be mentioned as an example. The performed analysis demonstrated that even when the worst initial conditions and possible uncertainties are considered, the fuel cladding remains intact. However, the analysis was performed assuming a fresh fuel assembly. In this study, an analysis of the fuel rod cladding behavior in the RBMK-1500 reactor following a bounding RIA is performed using the computational codes FEMAXI-6 and RELAP5. The analysis is extended by modeling an oxide layer (nodular corrosion) on the external surface cladding. An uncertainty and sensitivity analysis was performed using a method developed by the Society for Plant and Reactor Safety, employing the Software for Uncertainty and Sensitivity Analyses, in order to evaluate the effect of the oxide layer on the inside and outside fuel rod temperatures. The results of the thermo-mechanical analysis (stress, strain, and enthalpy) for a local oxide layer with a thickness of 70 μm show that despite the exceeded limit of allowed linear power density, the fuel rod is under acceptable safety conditions.

Analisis Termal-hidrolik Reaktor Cepat Berpendingin Gas (Gas Cooled Fast Reactor) Menggunakan Metode Runge Kutta

The Research of Gas Cooled Fast Reactor (GCFR) thermal-hydraulics analysis has been done. This reseacrh aim to solve fuel rod heat conduction equation by Runge Kutta method and to get thermal-hydraulics parameters such as coolant axial temperature distribution, pressure drops, convection heat transfer coefficient, and fuel rod radial temperature distribution. Heat transfer of the reactors was assumted steady state (time independent) then obtained coolant inlet tempertaure about 450 o C, outlet temperature about 474,752 o C and convection heat transfer coefficient about 2,5903 W/cm 2 o C. Pressure drop by friction was 0,16968 bar, pressure drop by form was 0,31292 bar, pressure drop by gravity was 0,20580 bar and total pressure drop was 0,68838 bar. While centerline fuel obtained the maximum temperature of fuel rod about 2720,33812 o C and the lowest fuel rod temperature at cladding surface about 488,8205 o C.

Analisis Termal-hidrolik Reaktor Air Bertekanan (Pressurized Water Reactor) Menggunakan Metode LU Faktorisasi

The Research about Pressurized Water Reactors (PWR) thermal-hydraulics analysis has been done. This reseacrh aim to solve fuel rod heat conduction equation by LU Factorization method and to get thermal-hydraulics parameters such as coolant axial temperature distribution, pressure drops, convection heat transfer coefficient, and fuel rod radial temperature distribution. Heat transfer of the reactors was assumted steady state (time independent) then obtained coolant inlet tempertaure about 300 o C, outlet temperature about 326,6 o C and convection heat transfer coefficient about 1,43 W/cm 2 o C. Pressure drop by friction was 0,423 bar, pressure drop by form was 0,51 bar, pressure drop by gravity was 0,25 bar and total pressure drop was 1,183 bar. While centerline fuel obtained the maximum temperature of fuel rod about 1.589,2 o C and the lowest fuel rod temperature at cladding surface about 360,9953 o C

Numerical study on hydrodynamic and thermal characteristics of spent fuel pool

In spent fuel pool (SFP), the characteristics of flow and temperature fields incurred by the decay heat from the spent fuel were explored by computational fluid dynamics (CFD) based on Navier-Stokes equations. For decay heat = 0.282–8.466 MW, the numerical results reveal that the height of the maximum fuel-rod temperature ranges between 2.1 and 3.0 m from the bottom of the fuel rod. The larger decay heat causes the location of the maximum fuel-rod temperature to be near the middle of the fuel-rod height. When the decay heat is close to 8.466 MW, the zirconium-water reaction could be incurred by the maximum fuel-rod temperature (>1089 K). To ensure a safe SFP, it is necessary for a nuclear power plant to monitor the fuel-rod temperature in this height range. The research result is a technical reference for evaluation of the thermal effect associated with the spent fuel rods stored in SFP.

Modeling the UO2 ex-AUC pellet process and predicting the fuel rod temperature distribution under steady-state operating condition

Modeling uranium dioxide pellet process from ammonium uranyl carbonate - derived uranium dioxide powder (UO2 ex-AUC powder) and predicting fuel rod temperature distribution were reported in the paper. Response surface methodology (RSM) and FRAPCON-4.0 code were used to model the process and to predict the fuel rod temperature under steady-state operating condition. Fuel rod design of AP-1000 designed by Westinghouse Electric Corporation, in these the pellet fabrication parameters are from the study, were input data for the code. The predictive data were suggested the relationship between the fabrication parameters of UO2 pellets and their temperature image in nuclear reactor.