Fission Gas Research Articles

Neutronics/thermal-hydraulics/fuel-performance coupling for light water reactors and its application to accident tolerant fuel

To accurately assess the in-pile performance of the accident tolerant fuel (ATF) over the whole life for the light water reactors (LWR), a full coupling of neutronics/thermal-hydraulics/fuel-performance is needed. In this paper, a multiphysics coupling code system named NECP-CLAMPERL is developed based on the high-fidelity neutronics code NECP-X, the sub-channel thermal-hydraulics code CTF, the finite-element fuel performance code NECP-CALF and the multiphysics object-oriented simulation environment (MOOSE). The code system is then applied to evaluate the performance of U3Si2 fuel, one of the most promising ATFs, by simulating a typical fuel assembly in pressurized water-cooled reactors (PWR). The quantities of interest including the reactivity, power distribution, fuel temperature, clad stress, fission gas release, plenum pressure, fuel radial displacement, gap width and PCMI are presented and analyzed in comparison with the traditional UO2 fuel. Future works and other potential applications are also identified.

An improved correlation for gas release from nitride fuels

An improved correlation for gas release from nitride fuels is elaborated. Introducing empirical activation energies for migration of fission gases in presence of solid fission products and oxide impurities, it becomes possible to better reproduce existing experimental data sets for gas release in sodium and helium bonded rods. The suggested approach may assist in resolving the previously poorly understood dispersion in measured gas release for identical irradiation conditions.

Improvement of the BISON U3Si2 modeling capabilities based on multiscale developments to modeling fission gas behavior

• The recommended models for analyzing U 3 Si 2 fuel behavior with BISON are provided. • Lower-length scale methods have been used to reduce uncertainty in fission gas modeling of U 3 Si 2 . • Comparisons to PIE measurements of U 3 Si 2 fueled rodlets are acceptable. Uranium silicide ( U 3 Si 2 ) is a concept explored as a potential alternative to UO 2 fuel used in light water reactors (LWRs) since it may improve accident tolerance and economics due to its higher thermal conductivity and increased uranium density. U 3 Si 2 has been previously used in research reactors in the form of dispersion fuel which operates at lower temperatures than commercial LWRs. The research reactor data illustrated that significant gaseous swelling occurs as the fuel burnup increases. Therefore, it is imperative to understand the fission gas behavior of U 3 Si 2 under higher temperature LWR operating conditions. In this work, molecular dynamics and phase-field modeling techniques are used to reduce the uncertainty in select modeling assumptions made in developing the fission gas behavior model for U 3 Si 2 in the BISON fuel performance code. To support the implementation of a U 3 Si 2 fission gas model in BISON, cluster dynamics simulations of irradiation enhanced Xe diffusion have been carried out. Similarly, MD simulations were used to predict the athermal contribution due to atomic mixing during ballistic damage cascades. By combining our results with literature DFT data for thermal equilibrium diffusion, Xe diffusivity has been described over a wide range of temperatures for in-reactor conditions. These lower length scale informed models are then utilized in the assessment of BISON U 3 Si 2 modeling capabilities by simulating the ATF-1 experiments irradiated in the Advanced Test Reactor (ATR). Sensitivity analysis (SA) and uncertainty quantification (UQ) are included as part of the assessment process to identify where further experiments and lower length scale modeling would be beneficial. The multiscale modeling approach utilized in this work can be applied to new fuel concepts being explored for both LWRs and advanced reactors (e.g., uranium nitride , uranium carbide).

Irradiation-enhanced diffusion and diffusion-limited creep in U3Si2

U 3 Si 2 is an advanced fuel candidate due to its relatively high fissile density and attractive thermal properties . Compared to standard UO 2 fuel, there are significant data gaps for the thermophysical and thermomechanical properties of U 3 Si 2 . Point defect concentrations and mobilities under irradiation govern a number of important fuel performance properties, such as creep and fission gas release. In this work, we utilized density functional theory (DFT) data to inform a cluster dynamics framework to predict point defect concentrations in U 3 Si 2 under irradiation. Molecular dynamics (MD) simulations were used to examine the contribution of atomic mixing during ballistic cascades to diffusion, as well as the diffusivity of U and Si at grain boundaries. These atomic scale models for diffusivity were then used to inform a creep model based on bulk (Nabarro-Herring) and grain boundary (Coble) diffusional creep , and climb-limited dislocation creep . The model compares well against available experimental data and has been implemented in the BISON fuel performance code. A demonstration case using simple power profiles has been carried out, showing that negligible creep occurs due to the low temperatures experienced by U 3 Si 2 in-reactor, a consequence of its high thermal conductivity .

Development of fission gas release model for MOX fuel pellets with treatment of heterogeneous microstructure

ABSTRACT This study develops a new fission gas release (FGR) model for mixed oxide (MOX) fuel pellets with fundamentally heterogeneous microstructures depending on the mixing methods of UO2 powder and PuO2 powder in fabrication process. The Japan Atomic Energy Agency developed a fuel performance modeling software FEMAXI-8 with an original FGR model supposing the microstructure to be homogeneous. The new FGR model explicitly treats the heterogeneity by computing the fission gas migration in two substructures independently. The model was applied to analyses of irradiation tests at the Halden reactor, in which two types of MOX fuels had different heterogeneity in their microstructure, while other fuel specifications were similar. The homogeneous model significantly underestimated FGR from the fuel with a remarkably heterogeneous microstructure. The new FGR model gave better calculations of the FGR behaviors of the remarkably heterogeneous fuel, namely the earlier onset of FGR and larger FGR at the end of irradiation than the relatively homogeneous fuel. The improved agreement with the experimental observation confirmed a certain validity of the developed model for estimating FGR from MOX fuels and studying heterogeneity effects.

Recent trends in metallic fast reactor fuels research

Metallic nuclear fuels have been extensively investigated for more than 60 years. Early studies noted that metallic fuels exhibit characteristic irradiation behaviors that set them apart from oxide fuels. By performing a critical review of historical data, the current gaps in the understanding of metallic U-Zr and U-Pu-Zr fuels (e.g., constituent redistribution, swelling and fission gas release) and recommendations for future research direction can be identified. This paper reviews and highlights the key aspects of the metallic fuel irradiation behaviors, with respect to the gaps in current understanding, based on the research performed in the past 3 years.

Thermodynamic predictions of CANDU fuel volatilization and fission product behaviour under severe accident conditions

A comprehensive thermodynamic investigation has been performed on irradiated uranium dioxide CANDU fuel under representative severe accident conditions. Two cases have been explored in this work: Case I) the fuel is in contact with varying combinations of hydrogen and steam (i.e., mainly reducing), and Case II) the fuel is in contact with varying combinations of air and steam (i.e., oxidizing). The system thermochemistry has been computed for a wide range of gas mixtures for both cases, fuel to gas ratios, temperatures, and hydrostatic pressures. The calculations predict fission gas speciation and fuel volatilization based on equilibrium thermodynamics, including the formation of several secondary phases, changes in UO2±x stoichiometry, fission product solubility in various phases, melting, and vaporization. The overall objective of this work was to inform source term analyses in the industry standard toolset SOURCE under severe accident conditions of a CANDU station.

Viscoplastic behavior of a porous polycrystal with similar pore and grain sizes: Application to nuclear MOX fuel materials

This study deals with the viscoplastic behavior of a porous polycrystal with pores and grains of similar sizes. Such a microstructure can be encountered in irradiated nuclear Mixed OXide (MOX) fuel materials. MIcronized MASter blend (MIMAS) MOX are multi-phase materials mainly composed of two or three phases depending on their fabrication process. One of these phases corresponds to plutonium-rich agglomerates which strongly evolve during irradiation. The large Pu-rich agglomerates become highly porous due to the accumulation of fission gases and to the apparition of irradiation bubbles. In a past study, Wojtacki et al. (2020) showed that pores distributed inside the Pu-rich clusters have a strong impact on the overall viscoplastic behavior of MOX fuel, when considering a purely isotropic behavior for the Pu-rich clusters. In the present study, the impact of pores similar in size to the surrounding anisotropic grains on the overall viscoplastic behavior is studied in details through numerical full-field simulations. A crystal plasticity model recently developed by Portelette et al. (2018) is used to describe the anisotropic behavior of the polycrystalline matrix with dislocation glide mechanisms. Three-dimensional full-field simulations are performed by a method based on Fast Fourier Transforms (FFT) to compare the behavior of porous materials with that of dense materials. These simulations show that, in the case of spherical pores, their relative size with respect to that of the grains plays a minor role in the overall viscoplastic behavior. However, in the case of polyhedral pores, the relative size effect is more pronounced. With fixed porosity, decreasing the relative size of the cavities with respect to the size of grains leads to a softening of the material and a decrease of the viscoplastic flow stress.

Interaction of krypton and xenon with sodium and activated charcoal: Identification and modeling using gas chromatography and density functional theory

For the online detection of failed fuel pin in Prototype Fast Breeder Reactor (PFBR), Failed Fuel Localization Module (FFLM) facility is designed to probe the sodium (Na) coolant. Fission Gas Sparging Module (FGSM) is retrofitted with FFLM to localize failed fuel subassembly during the dry rupture of the fuel pin. FGSM is used to purge gaseous fission products (krypton (Kr) and xenon (Xe)) from Na, using argon (Ar). In this work, the efficiency of separating inactive Kr and Xe noble gases with Ar purge out of liquid Na, by their collection onto activated charcoal at cryogenic temperatures and the consequent release at high temperatures into the gas samplers was studied using gas chromatography (GC). Studies of the recovery of Xe/Kr separately with and without Na matrix in the FGSM under both static and dynamic conditions were undertaken by sequentially collecting these probe gases along with excess of Ar in different gas samplers which were then estimated using GC. Density Functional Theory (DFT) computations were performed on model Na clusters surrounding Kr/Xe atoms and also on pyrene molecule (a prototype for simulating the spaces within charcoal), trapping Kr and Xe atoms. Experimental observations showed a good correlation with the results obtained from Quantum Theory of Atoms In Molecules (QTAIM), Natural Bond Orbital (NBO), Energy Decomposition (ED), Non-Covalent Interaction (NCI) and Electrostatic Potential (ESP) mapping analyses. Calculated thermochemical parameters affirm the binding of Xe with Na and charcoal to be marginally stronger than that of Kr, consistent with the experimental observations and speculations based on their polarizabilities.

Microstructural and Micro-Chemical Evolutions in Irradiated UCO Fuel Kernels of AGR-1 and AGR-2 TRISO Fuel Particles

AGR-1 and AGR-2 tristructural-isotropic (TRISO) fuel particles were fabricated using slightly different fuel kernel chemical compositions, modified fabrication processes, different fuel kernel diameters, and changed 235U enrichments. Extensive microstructural and analytical characterizations were conducted to correlate those differences with the fuel kernels’ responses to neutron irradiations in terms of irradiated fuel microstructure, fission products’ chemical and physical states, and fission gas bubble evolutions. The studies used state-of-the-art transmission electron microscopy (TEM) equipped with energy-dispersive x-ray spectroscopy (EDS) via four silicon solid-state detectors with super sensitivity and rapid speed. The TEM specimens were prepared from selected AGR-1 and AGR-2 irradiated fuel kernels exposed to safety testing after irradiation. The particles were chosen in order to create representative irradiation conditions with fuel burnup in the range of 10.8 to 18.6% fissions per initial metal atom (FIMA) and time-average volume-average temperatures varying from 1070 to 1287°C. The 235U enrichment was 19.74 wt.% and 14.03 wt.% for the AGR-1 and AGR-2 fuel kernels, respectively. The TEM results showed significant microstructural reconstructions in the irradiated fuel kernels from both the AGR-1 and AGR-2 fuels. There are four major phases: fuel matrix of UO2 and UC, U2RuC2, and UMoC2—in the irradiated AGR-2 fuel kernel. Zr and Nd form a solid solution in the UC phase. The UMoC2 phase often features a detectable concentration of Tc. Pd was mainly found to be located in the buffer layer or associated with fission gas bubbles within the UMoC2 phase. EDS maps qualitatively show that rare-earth fission products (Nd et al.) preferentially reside in the UO2 phase. In contrast, in the irradiated AGR-1 fuel kernel, no U2RuC2 or UMoC2 precipitates were positively identified. Instead, there was a high number of rod-shaped precipitates enriched with Ru, Tc, Rh, and Pd observed in the fuel kernel center and edge zone. The differences in irradiated fuel kernel microstructural and micro-chemical evolution when comparing AGR-1 and AGR-2 TRISO fuel particles may result from a combination of irradiation temperature, fuel geometry, and chemical composition. However, irradiation temperature probably plays a more deterministic role. Limited electron energy-loss spectroscopy (EELS) characterizations of the AGR-2 fuel kernel show almost no carbon in the UO2 phase, but a small fraction of oxygen was detected in the UC/UMoC2 phase.

Release Rates of Short-lived Fission Gases from Modern Spherical Fuel Elements with TRISO-coated Particles

Measurement and prediction of fission product release is important in design, licensing and operation of HTRs. Latest development status is near zero particle manufacturing defects and in-pile failures combined with very low matrix contamination levels. In-pile Release over Birth rate measurements will be described and evaluated. Quantitative predictive analyses of short-lived gas release will be provided for MTR irradiation testing with a focus on the most recent experiment with Chinese fuel from HTR-PM production. For irradiation tests with no single particle failure, models describing particle kernel release are not applicable; instead modeling of matrix contamination release is essential. Good agreement between post calculations and in-line R/B measurements has been achieved.

Failure Fraction Calculation of the TRISO-Coated Particle Using X-Ray Computed Tomography

The geometrical shape of the TRISO-coated particle is closely related to its performance and safety. In this paper, models were set up to study the failure fraction of TRISO particle, considering the real asphericity induced by manufacturing uncertainties. TRISO is simplified as a pressure vessel model, and micro X-ray CT was employed to detect the real geometrical shape. Key geometrical parameters, thickness and volume of the real particle, were then obtained with the 3D measurement method and input into PANAMA code (a German code for fuel performance simulation). Release fraction of fission gas and failure fraction of the TRISO-coated particle were revised with the aforementioned parameters with more accuracy and compared with those of the spherical particle. Obvious increment of failure fraction of the particle is found, which may contribute to the release of fission products.

3-Dimensional Multiphysics Modeling of Behaviors of Pressurized Water Reactor Fuel Rods With Missing Pellet Surface

A 3-dimensional (3D) fuel performance analysis program, able to simulate normal operating conditions and accident conditions for PWR fuel behaviors, was developed based on the Multiphysics Object-Oriented Simulation Environment (MOOSE) finite-element framework. By taking fission products swelling, densification and expansion of pellet, thermal and irradiation creep, gap heat transfer, fission gas release, and cladding crack propagation into consideration, detailed fuel behaviors have been simulated in a multiphysics coupling way. Local defects in fuel pellet caused during manufacturing and filling processes known as the missing pellet surface (MPS) can cause abnormal stress distribution of the cladding and it could even lead to cladding failure. Taking Stress Corrosion Cracking (SCC) phenomenon into consideration, a simulation of PWR fuel rodlet that consists of a pellet with an MPS defect and an intact pellet was conducted. The fuel rod has experienced with sorts of events, including normal operating conditions and a high-power ramp event. The simulation results indicated that: 1) The MPS defect affects the temperature and displacement distribution in the vicinity of the MPS defect. When the pellets are in contact with the cladding, the inner surface of the cladding presents a large tensile hoop stress, which accelerates the crack propagation. 2) During the ramp event, the crack propagation rate was higher than that under normal condition and crack length expanded by about 0.1 µm.

Development of fuel rod behavior analysis code and its application to supercritical CO2 cooled nuclear reactor

To simulate the thermal–mechanical performance of a Supercritical CO2-cooled reactor fuel during the long-term normal operation, material models and new thermal and mechanical models are implemented in FROBA code to develop FROBA-SCO2. By calculation results comparison of the new models and ABAQUS, KAIST code and FRAPCON code, the reliability of new models were validated. The results show that with the effects of densification and burnup, the size and the temperature drop of the gas gap firstly increase and then decrease. The stress and the train of the cladding are affected by the pressure drop from both sides and change slightly. The fission gas release fraction is at a low level and the gas pressure increases slightly. The cladding corrosion behavior increases with the increment of temperature. The result indicated that the fuel element was safe enough during most time of the normal operation. The models established in this article calculates the fuel rod behaviors efficiently.

Initial demonstration of automated fuel performance modeling with 1977 EBR-II metallic fuel pins using BISON code with FIPD and IMIS databases

Using the BISON fuel performance code, simulations were conducted using an automated process to read initial and operating conditions from the Fuels Irradiation and Physics Database (FIPD) and Integral Fast Reactor materials information system (IMIS) database, which contains metallic fuel data from the Experimental Breeder Reactor-II (EBR-II). This work demonstrates use of an integrated framework to access the vast majority of EBR-II experimental fuel pin data to support rapid development of fuel performance models for next-generation metallic fuel systems. With this capability, validation for fuel qualification can be performed rapidly. Between IMIS and FIPD, there is enough information to conduct 1977 unique EBR-II metallic fuel pin histories from 24 different experiments, at varying levels of detail between the two databases. Each of these histories includes a high-resolution power history, flux history, coolant channel flow rates, and coolant channel temperatures. Fission gas release (FGR), cumulative damage fraction (CDF), fuel axial swelling, cladding profilometry, and burnup were all simulated in BISON. The results were compared to post-irradiation examination (PIE) results for the initial demonstration of automated BISON modeling. BISON simulations conducted with IMIS and FIPD were in rough agreement with PIE measurements and calculations. Cladding profilometry, FGR, and fuel axial swelling were found to be in rough agreement with PIE measurements, depending on the physics used within the BISON input files. The mechanical contact solver chosen was found to significantly impact axial fuel swelling and cladding strain predictions. CDF values were assessed to see whether pin failure may have been predicted (CDF ≥ 1). This work suggests that continued development of an automated tool for BISON should focus on inclusion of the Fast Flux Test Facility (FFTF) experimental data for a larger database for metallic fuel, improved physical models to better capture fuel performance, such as fuel-cladding interactions, and a more detailed comparison with available PIE data to further the BISON model development.

Lead-Cooled Fast Reactor Annular UN Fuel Design and Development of Performance Analysis Program

A kind of annular uranium nitride (UN) fuel suitable for lead-cooled fast reactor applications has been designed in this study. The design is directly targeting two main issues of UN fuel: severe swelling and thermal decomposition of UN fuel at high temperatures. A performance analysis program based on FORTRAN programming language has been developed for UN fuel in fast reactors. The program contains heat transfer, fuel stress-strain analysis, cladding stress-strain analysis, fission gas release and fuel-cladding mechanical interaction (FCMI) modules, etc. Extensive code verification has been performed by comparing simulation results obtained with the code and those obtained via the COMSOL Multiphysics platform. Preliminary code validation has been conducted as well by comparing code simulation results with experimental data. The results showed that this program could predict the fuel temperature, stress-strain, and displacement of UN fuel during reactor operation with a reasonable accuracy.

Microstructural Changes and Chemical Analysis of Fission Products in Irradiated Uranium-7 wt.% Molybdenum Metallic Fuel Using Atom Probe Tomography

Understanding the microstructural and phase changes occurring during irradiation and their impact on metallic fuel behavior is integral to research and development of nuclear fuel programs. This paper reports systematic analysis of as-fabricated and irradiated low-enriched U-Mo (uranium-molybdenum metal alloy) fuel using atom probe tomography (APT). This study is carried out on U-7 wt.% Mo fuel particles coated with a ZrN layer contained within an Al matrix during irradiation. The dispersion fuel plates from which the fuel samples were extracted are irradiated at Belgian Nuclear Research Centre (SCK CEN) with burn-up of 52% and 66% in the framework of the SELENIUM (Surface Engineering of Low ENrIched Uranium-Molybdenum) project. The APT studies on U-Mo particles from as-fabricated fuel plates enriched to 19.8% revealed predominantly γ-phase U-Mo, along with a network of the cell boundary decorated with α-U, γ’-U2Mo, and UC precipitates along the grain boundaries. The corresponding APT characterization of irradiated fuel samples showed formation of fission gas bubbles enriched with solid fission products. The intermediate burnup sample showed a uniform distribution of the typical bubble superlattice with a radius of 2 nm arranged in a regular lattice, while the high burnup sample showed a non-uniform distribution of bubbles in grain-refined regions. There was no evidence of remnant α-U, γ’-U2Mo, and UC phases in the irradiated U-7 wt.% Mo samples.

Preliminary analysis on the thermal-mechanical behavior of dispersed plate-type fuel under reactivity insertion accident

In this study, a modified two-dimensional geometry model was employed comprised of metal cladding, a metal matrix, and dispersed fuel particles. And an analysis method was established to simulate the thermal–mechanical behavior of dispersed plate-type fuel under the Reactivity Insertion Accident (RIA). Key fuel properties including thermal performance, mechanical behaviors and irradiation effects were considered. All these models were coded in the ABAQUS subroutines and solved. The influences of heat transfer deterioration, fission gas and burnup were investigated. Significant parameters to describe the performance of the fuel element, such as temperature, von Mises stress and strain profile, were evaluated. The results indicate that the deterioration of the heat transfer in the flow channel and the deepening burnup increase stress and strain of the dispersed plate-type fuel. The fission gas causes the crack of the fuel particles.

Study on the relationship between fuel fragmentation during a LOCA and pellet microstructure

ABSTRACT To systematically examine the relationship between the fuel fragmentation during a loss-of-coolant accident (LOCA) and the pellet microstructure, heating tests were conducted on micro fuel samples obtained from various radial positions of high-burnup fuel. The extent of the fuel fragmentation was evaluated via fission gas release rate and post-test microscopies. The results provide direct evidence for the contribution of the high-burnup structure (HBS) on the fine fuel fragmentation during a LOCA. The paper also demonstrates that the samples with a local burnup of lower than 72 GWd/tU did not fragment severely, which is in good agreement with the previously proposed burnup threshold. Heating tests incorporating stepwise depressurization from high pressure were also conducted to examine the evolution of fission gas release throughout the depressurization. The suppressing effect of hydrostatic pressure on fuel fragmentation is further confirmed. Additionally, the paper also provides results that suggest that most of the fine fragmentation at the HBS occurs during depressurization from 20 to 2 MPa.

Effect of gap design pressure on the LWR fuel rods lifetime

Abstract The fuel burn-up rate has been raised in recent years to improve the efficiency of nuclear LWRs (light water reactors). Therefore, surveying and estimating changes in fuel properties and structural materials during radiation exposure is of paramount importance. In the present study, the researchers focused on analyzing the role of LWR fuel rod initial gap pressure (initial gas pressure when a fuel rod is fabricated) on the rod’s thermal and mechanical performance. FRAPCON-4.0 steady-state fuel performance code was used to simulate the effect of initial gap pressure on the behavior of a specific BWR-type fuel rod that was irradiated under the HALDEN research program. This fuel rod is similar to commercial BWR fuel rods in all respects, except that the research reactors have a height limit. The important fuel design criteria, such as the centerline temperature, effective stresses, total released fission gas to the fuel rod’s void volumes, and the cladding strains, were included in the analysis. According to the present study, a potential initial gap pressure range could be suggested to increase fuel rods’ lifetime by improving the safety criteria margins, especially fuel centerline temperature and the released amount of gaseous fission products. As we know, lower fuel temperature leads to having a reactor with a higher power density and, consequently, a maximum fuel burn-up rate, which can affect the economy and safety of nuclear power plants.