Uranium dioxide properties for LWR fuel rods

Uranium dioxide pellets have become the most important nuclear fuel, and will remain so far a long time, with the fissile isotope ²³⁵U being replaced by PuO₂ additions. This does not significantly change the pellet properties.Uranium dioxide properties affect fuel rod performance more than previously anticipated, because UO₂ pellets show a distinct response to irradiation, and because of mechanical and chemical interaction with cladding. Here elastic and plastic behaviour, fracturing, irradiation densification, and dimensional behaviour under steady and power cycling conditions are mainly covered.     

Mathematical modeling and analysis of fuel rods

This paper presents a comprehensive method of analysis for nuclear fuel rods by first developing appropriate constitutive relations for the fuel and the clad materials that take into account elastic-plastic behavior, primary as well as secondary creep, mechanical densification (hot pressing), fuel swelling, fuel cracking and healing, and fuel redistribution. The geometric problem is treated on the basis of a two-dimensional, axisymmetric, and plane model within the framework of the finite element method. General power histories of the transient, cyclic, and steady-state types are considered. Analysis results uncovering the importance of certain deformation processes, which heretofore have been neglected, and comparisons with experiments are given.     

The DIMCO system for mechanical design of fuel rods

This study presents the system developed at JEN for one- and two-dimensional mechanical analysis of nuclear fuels. The mathematical and numerical bases are described, as well as different models representing phenomena such as cracking, swelling, etc. Numerical results of several test cases are presented: (a) checking of one- and two-dimensional analyses, (b) applicability margin, (c) interaction of effects; and (d) influence of loading histories.     

Fuel oxidation and thermal conductivity model for operating defective fuel rods

A model has been developed to describe the fuel oxidation behaviour, and its influence on the fuel thermal conductivity, in operating defective nuclear fuel rods. The fuel-oxidation model is derived from adsorption theory and considers the influence of the high-pressure environment that results from coolant entry into the fuel-to-clad gap. This model is in agreement with the fuel-oxidation kinetics observed in high-temperature annealing experiments conducted at 1473-1623 K in steam over a range of pressure from 0.001 to 0.1 MPa. Using a Freundlich adsorption isotherm, the current model is also consistent with recent experiments conducted at a higher pressure of 7 MPa. The model also considers radiolytic effects as a consequence of fission fragment bombardment in the fuel-to-clad gap. This treatment suggests that radiolysis-assisted oxidation is insignificant in operating defective rods (as compared to thermal effects), as supported by limited in-reactor data. The effects of diffusion of the interstitial oxygen ions in the solid in the operating rod is further discussed.     

Calculational procedure for determining creep collapse of LWR fuel rods

Observed collapses in pressurized water reactor fuel rods have been attributed to the radiation enhanced creep of Zircaloy cladding into regions where separations in the fuel pellet stack have occurred. A computer code, COLAPX, has been written to determine the growth of ovality and the ultimate collapse of fuel rod cladding under reactor operating conditions. This paper describes the theoretical bases of this code, the finite element formulation used, the constitutive relations between the displacement fields and the element forces, and the radiation, temperature and stress dependent material model for creep of Zircaloy tubing. Comparisons of the creep rate predictions and of the ovality predictions with data from irradiated tubes and fuel cladding are presented.     

A mathematical model for investigation of the rewetting of fuel rods

The computer model ZETHYF simulating the reflood phase after a loss-of-coolant accident with emphasis on the investigation of coolant channel is described. The thermal behaviour of the fuel rod is modeled based on a detailed representation of the heat transfer mechanisms and a moving mesh around the quench front. The flow conditions in the coolant channel are simulated as a one-dimensional transient one- or two-phase flow.     

TRISO-fuel element thermo-mechanical performance modeling for the hybrid LIFE engine with Pu fuel blanket

A TRISO-coated fuel thermo-mechanical performance study is performed for the fusion-fission hybrid Laser Inertial Fusion Engine (LIFE) to test the viability of TRISO particles to achieve ultra-high burn-up of Pu or transuranic spent nuclear fuel blankets. Our methodology includes full elastic anisotropy, time and temperature varying material properties, and multilayer capabilities. In order to achieve fast fluences up to 30 × 10²⁵ n m⁻² (E > 0.18 MeV), judicious extrapolations across several orders of magnitude of existing material databases have been carried out. The results of our study indicate that failure of the pyrolytic carbon (PyC) layers occurs within the first 2 years of operation. The particles then behave as a single-SiC-layer particle and the SiC layer maintains reasonably-low tensile stresses until the end-of-life. It is also found that the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Conversely, varying the geometry of the TRISO-coated fuel particles results in little differences in terms of fuel performance.     

Transmission computer microtomography for nondestructive monitoring of fuel and control rods in a reactor

A method of x-ray transmission computer microtomography has been developed for solving the problems of monitoring the quality of fuel elements and control rods in nuclear reactors: the geometric resolution of a defect is several microns. The solution of the problem of nondestructive monitoring of such objects has made it possible to perform a wide range of investigations of the technical characteristics of definite x-ray detectors, organize principles of scanning, based on the principles of laser interferometry and the design of data processing systems, that are different from those of conventional systems. The investigations performed have made it possible to implement computer-aided design for problem-oriented computer microtomographs and for the instrumentational implementation of an experimental variant of such a device. Investigations performed on specially fabricated test samples with calibrated defects have demonstrated that the approach to the design of such setups for nondestructive monitoring of objects for nuclear power generation is correct and that this is a promising direction, 8 figures, 9 references. All-Russia Scientific-Research Institute of Automatic Machine Engineering. Translated from Atomnaya énergiya, Vol. 88, No. 2, pp. 125–137, February, 2000.     

Some analytical and numerical results for cladded fuel rods subjected to thermo-irradiation induced creep

Stresses and velocities are analyzed for a hollow cylinder of fuel encased in metallic cladding and subjected to high temperature and high neutron flux fields. The material is represented by a compressible nonlinear thermo-irradiation viscoelastic model. A stress function for axisymmetric plane strain is introduced, and the problem essentially reduces to solving a nonlinear ordinary differential equation in the interfacial contact pressure. Some analytical results are obtained for the case of material properties independent of position. For the case of temperature, flux and thus material properties dependent on position, an approximate formulation is employed whereby the cylinder is divided into discrete rings with constant properties.     

A quantitative estimate on the heat transfer in cylindrical fuel rods to account for flux depression inside fuel

In a nuclear reactor, the power is limited by thermal rather than by nuclear considerations. The reactor core must be operated at a power level that the temperatures of the fuel and cladding anywhere in the core must not exceed safety limits to avoid damages in the fuel elements.Heat transfer from fuel pins can be calculated analytically by using a flat power density in the fuel pin. In actual practice, the neutron flux distribution inside fuel pins results in a smaller effective distance for the heat to be transported to the coolant. This inherent phenomenon gives rise to a heat transfer benefit in fuel pin temperatures.In this research, a quantitative estimate for transferring heat from cylindrical fuel rods is accomplished by considering a non-uniform neutron flux, which leads to a flux depression factor. This, in turn, alters the temperature inside the fuel pin. A theoretical relationship combining the flux depression factor and a ratio of temperature gradients for uniform and non-uniform is derived, and a computational program, based on finite volume method and energy balance, is developed to validate the considered approximation.     

Power ramp induced iodine and cesium redistribution in LWR fuel rods

Volatile fission product migration in LWR fuel rods which are power ramped above a certain threshold beyond the envelope of their previous power history, plays an important role in stress corrosion cracking of Zircaloy. This may cause fuel rods to fail already at stresses below the yield strength. In the HFR, Petten, many power ramp experiments have been performed with subsequent examination of the ramped rods for fission product distribution. This study describes the measurement of iodine and cesium distribution using γ-spectroscopy of I-131 and Cs-137. An evaluation method is presented which makes the determination of absolute amounts of I/Cs feasible. It is shown that a threshold for I/Cs redistribution exists beyond which it depends strongly on local fuel rod power and fuel type.     

Optimisation of hot isostatic pressing bonded SS/SS joints conditions for ITER blanket shield

In the engineering design activity of international thermonuclear experimental reactor (ITER), stainless steels are being considered as candidates materials for several module type structures. Hot isostatic pressing (HIP) technique is expected for the fabrication of these modules. Stainless steel powders are simultaneously consolidated as mono-material block or/and joined in bi-material module.This paper reviews the manufacturing stages, non-destructive examination and the developments of the HIP bonded joints of 316L SS (powder and solid) for application to the ITER shield blanket.It is well known that the powder surface oxidation negatively influences the impact toughness of raw material and joints consolidated by this way. In order to get acceptable mechanical properties of materials, a study on the effect of reducing the powder oxygen content has been launched. To evaluate susceptibility to the oxygen content of HIPed joint specimens, tensile and toughness tests have been performed.From this study, optimal conditions of HIP were fitted and the influence of oxygen was mastered to obtain good mechanical properties of the consolidated powder material as well as for HIPed junction.     

Temperature distribution in mixed ThO2–UO2 fuel rods located in blanket of an inertial fusion energy breeder

Investigations of neutronic analysis and temperature distribution in fuel rods located in a blanket driven ICF (Inertial Confinement Fusion) have been performed for various mixed fuels and coolants under a first wall load of 5 MW/m². The fuel rods containing ThO₂ and UO₂ mixed by various mixing methods for achieving a flat fission power density are replaced in the blanket and cooled with different coolants; natural lithium, flibe, eutectic lithium and helium for the nuclear heat transfer. It is assumed that surface temperature of the fuel rod increases linearly from 500 °C (at top) to 700 °C (at bottom) during cooling fuel zone. Neutronic and temperature distribution calculations have been performed by MCNP4B Code and HEATING7, respectively. In the blanket fueled with pure UO₂ and cooled with helium, M (fusion energy multiplication ratio) increases to 3.9 due to uranium having higher fission cross-section than thorium. The high fission energy released in this blanket, therefore, causes proportionally increasing of temperature in the fuel rods to 823 °C. However, the M is 2.00 in the blanket fueled with pure ThO₂ and cooled with eutectic lithium because of more capture reaction than fission reaction. Maximum and minumum values of TBR (tritium breeding ratio) being one of main neutronic paremeters for a fusion reactor are 1.07 and 1.45 in the helium and the natural lithium coolant blanket, respectively. These consequences bring out that the investigated reactor can produce substantial electricity in situ during breeding fissile fuel and can be self-sufficient in the tritium required for the DT fusion driver in all cases of mixed fuels and coolant types. Quasi-constant fission power density profiles in FFB (fissile fuel breeding) zone are obtained by parabolically increasing mixture fraction of UO₂ in radial and axial directions for all coolant types. Such as, in the helium coolant blanket and the case of PMF (parabolically mixed fuel), Γ (peek-to-average fission power density ratio) of the blanket is reduced to 1.1, and the maximum temperatures of the fuel rods in radial direction of the FFB zone are also quasi-constant. At the same time, in the case of PMF, for all coolant types, the temperature profiles in the radial direction of the fuel rods rise proportionally with surface temperature from the top to the bottom of fuel rods in the axial direction. In other words, for each radial temperature profile in the axial direction, temperature differences between centerline and surface of the fuel rods are quasi-constant. According to the coolant types, these temperature diffences vary between 30 and 45 °C.     

Modeling of gap conductance for LWR fuel rods applied in the BISON code

ABSTRACT

A new gap conductance model is proposed in this study as a combination of Toptan’s model and the Ross-Stoute model. A variance-based sensitivity analysis is performed to understand how simulation results depend on all input parameters of the proposed model. Additionally, new modeling options (e.g. fill gas thermal conductivity, temperature jump distance, thermal accommodation coefficient, etc.) are added into the nuclear fuel performance code, BISON. The need for further investigation of the gap heat transfer between fuel and cladding in BISON motivated this study to evaluate its impact on the code’s predictions. New gap conductance modeling is proposed. A series of integral-effects validation tests is performed: (1) to demonstrate the impact of the proposed model on the code’s fuel temperature predictions at the beginning of life and through the reactor’s life; (2) to ensure that the proposed model is capable of accurately modeling gap heat transfer characteristics in real-world problems; and (3) to investigate the impact of the estimation of fission gas release on the fuel temperature predictions with the proposed model. The results indicate that the proposed gap conductance model improves BISON’s predictions.     

A hot cell for the reprocessing of spent fuel samples

The hot cell for the reprocessing of spent fuel samples is suggested to be erected in the process hall of the radiochemical laboratory at Inchas. The design permits for safe handling of spent fuel samples with radioactivity level up to 10,000 MeV-Ci. The hot cell consists of an air-tight stainless steel box, 2 × 1.5 × 1.75 m³, surrounded by suitable biological shielding, and equipped with masterslave manipulators, lead glass window, and a number of gloves. The box is connected with a solid-waste disposal mechanism, and a feeding shute fo rtransferring the spent fuel samples into the hot cell. The cell is provided with a sliding door through which the stainless steel box can be easily withdrawn for maintenance or replacement. The ventilation of the hot cell ensures non-radioactivity release in the surrounding zones.     

Isotope correlation technique for Pu mass analysis of PWR UO2 spent fuel rods

A simple and fast method of nuclear material accountancy of pressurized water reactor (PWR) UO₂ spent fuel rods for safeguards application was developed utilizing the isotope correlation between the amounts of ¹³⁷Cs and total Pu. To this end, the following steps were taken: (1) as much destructive analysis (DA) data as possible for segments taken from a PWR UO₂ spent fuel rod were aggregated from publicly available data sources; (2) the DA data were corrected so as to have the same cooling time (i.e., CT = 0 y) and analyzed for outliers; (3) an equation converting the ¹³⁷Cs amount to the Pu amount was obtained by regression analysis with logarithmic curve fitting; and (4) the error in determining the Pu amount was evaluated for the imposition of a limit on the range of burnup (BU) or initial enrichment (IE). It was found that the averaged % error in calibration was determined to be 3.88% ± 2.68% (= mean ± 1 standard deviation) for the BU range over 30 GWd/tU and falling with increasing BU range. On the other hand, there was no benefit in applying the limit of the IE range. Lastly, the Pu-mass difference between various methods was compared and it was found that the difference can be incurred up to 11.4%, according to the choice of method. In conclusion, the proposed isotope correlation technique could be used for input material accountancy with reasonable uncertainty.     

The anisotropy of the high temperature deformation of Zircaloy-cladding tubes for nuclear fuel rods

The anisotropy of the high temperature deformation of Zircaloy-4 cladding tubes for nuclear fuel rods for pressurized water reactors has been investigated. The axial and tangential components of the deformation of internally pressurized tube samples during closed end creep rupture tests in air at 800°C have been measured. An axial contraction of the tube sample is observed. Using Hill's theory of plasticity the axial strain can be described by anisotropy coefficients which depend on the texture of the tube material. The anisotropy coefficients are quantitatively related to the orientation distribution of the basal poles in the radial/tangential plane of the tube sample. For the typical texture of Zircaloy cladding tubes of nuclear fuel rods for pressurized water reactors, an axial contraction has to be expected under the biaxial stress conditions applied.     

FRAP-T6: A computer code for the transient analysis of oxide fuel rods

The FRAP-T6 computer code was developed to model the transient performance of light water reactor fuel rods during reactor transients ranging from mild operational transients to large break loss-of-coolant accidents. The code models all of the thermal, structural, and chemical phenomena needed for the complete evaluation of light water reactor fuel rod performance. The code was developed using rigorous quality assurance procedures and a large assessment data base. The results of assessment show that the code accurately models the response of light water reactor fuel rods.     

高燃耗下M5合金包壳燃料棒稳态和瞬态性能分析

修改并验证了分析程序FEMAXI-IVM,增加了程序的适用范围。对采用M5合金包壳的FA300-4高性能燃料组件中的燃料棒在稳态和瞬态运行工况下的燃料性能进行了分析。结果表明,此种燃料棒在稳态和瞬态工况下都能保持其完整性,能保证反应堆的安全运行。