Fission Gas Release Research Articles

Analysis of pellet-to-cladding gap closure for a metallic micro-cell pellet under normal operating conditions

To reduce fission gas release under both accident conditions and normal operation, Korea Atomic Energy Research Institute has developed a metallic micro-cell pellet following the 2011 Fukushima accident. One concern, however, about the metallic micro-cell pellet is that due to the presence of metallic wall that reduces both temperature and gas release, its capability of higher gas retention could result in increased gaseous swelling, leading to earlier pellet-to-cladding gap closure than the conventional UO 2 pellet. To analyze and compare the gap closure in two fuel rods in which two kinds of pellets, UO 2 and metallic micro-cell, are loaded into each of the two identical Zr claddings, we developed a model to estimate gaseous swelling in the two pellets. Then considering mainly different gaseous swelling and thermal expansion by different temperature profiles in the two pellets, we compared the pellet-to-cladding gaps as a function of burnup for the case where they would be irradiated with the same constant linear power of 30 kW/m up to 60 MWd/kgU. Due to both reduced gaseous swelling and lower thermal expansion, gap closure in the fuel rod loaded with the metallic micro-cell pellet is expected to be delayed by 6–16 MWd/kgU compared with the one loaded with the UO 2 pellet, implying pellet-cladding mechanical interaction occurrence would also be postponed by the same degree. We note that the result of this approach is preliminary until confirmed by measured data from metallic micro-cell pellets irradiated up to high burnup. • Higher gas-retention capability in the metallic micro-cell pellet could lead to earlier pellet-to-cladding gap closure than the UO 2 pellet. • Despite the higher gas-retention capability, gaseous swelling in the metallic micro-cell pellet would be reduced by its lower temperature. • Pellet-to-cladding gap closure in the metallic micro-cell pellet could be delayed by 6–16 MWd/kgU compared with the UO 2 pellet.

25-Pin metallic fuel performance benchmark case based on the EBR-II X430 experiments series

A metallic fuel benchmark case was developed for the fuel performance code BISON based on 25 uranium-zirconium and uranium-plutonium-zirconium pins of the Experimental Breeder Reactor II X430 experiment series. Results of the benchmarks were compared with measurements and calculations made at the time of the experiment as well as subsequent measurements reported in 2019. The comparisons were used to quantify the accuracy of the BISON predictions and to identify patterns in the BISON differences. BISON predicted burnup , plenum pressure, and fission gas release fractions accurately. BISON temperature predictions were somewhat cooler than the temperatures determined at the time of the experiment but appeared to be reasonably accurate considering uncertainties in the legacy temperature calculations, uncertainties in the legacy linear heat rate calculations, and the high sensitivities of the BISON-predicted temperatures to BISON inputs. Fuel axial elongation predictions had errors correlated to fuel composition; BISON tended to underpredict the elongation of binary fuels and overpredict the elongation of ternary fuels. BISON cladding radial dilation predictions were also significantly lower than legacy PIE measurements. Recommendations were made to improve the BISON fuel gaseous swelling model to account for fuel composition, to add additional capabilities to the coolant channel temperature model to ease benchmark development, and to continue developing benchmark cases based on a wide range of experiments in several reactors. Once a wide array of benchmarks is developed, an attempt can be made to enhance or calibrate BISON models to improve the cladding dilation predictions.

On the intra-granular behaviour of a cocktail of inert gases in oxide nuclear fuel: Methodological recommendation for accelerated experimental investigation

Besides recent progresses in the physics-based modelling of fission gas and helium behaviour, the scarcity of experimental data concerning their combined behaviour (i.e., cocktail) hinders further model developments. For this reason, in this work, we propose a modelling methodology aimed at providing recommendations for accelerated experimental investigations. By exploring a wide range of annealing temperatures and cocktail compositions with a physics-based modelling approach we identify the most interesting conditions to be targeted by future experiments. To corroborate the recommendations arising from the proposed methodology, we include a sensitivity analysis quantifying the impact of the model parameters on fission gas and helium release, in conditions representative of high and low burnup.

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.

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.

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.

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.

EBSD studies on microstructure and crystallographic orientation of UO2-Mo composite fuels

The microstructure of the fuel pellet plays an essential role in fission gas buildup and release and is critical for the safe and continued operation of nuclear power stations. Structural analysis of uranium dioxide (UO2)–molybdenum (Mo) composite fuel pellets prepared at a range of sintering temperatures from 1300 to 1800 °C was performed. Mo micro and nanoparticles were used in making the composite pellets. A systematic investigation into the influence of processing parameters during Spark Plasma Sintering (SPS) of the pellets on the microstructure, texture, grain size, and grain boundary characters of UO2-Mo is presented. UO2-Mo composite show significant differences in the fraction of general boundaries and also special/coincident site lattice (CSL) boundaries. EBSD orientation maps demonstrated that <111> texturing was observed in the pellets fabricated at 1500 °C. The experimental investigations suggest that UO2-Mo composite pellets have favorable microstructural features compared to the UO2 pellet.

Fuel performance analysis of reduced moderated boiling water reactor for transuranic waste incineration

Reduced moderated boiling water reactors (BWR) could serve as near-term alternative to sodium fast reactors for the application of transuranic waste incineration. One such concept currently proposed by Hitachi is the Resource-Renewable Boiling Water Reactor for Transuranic Burning (RBWR-TB2) that utilizes the proven advanced BWR (ABWR) technology. FRAPCON-3.5 EP with its capability to model the physical phenomena at high temperature and high burnup conditions is used to compare the fuel performance of RBWR-TB2 and ABWR fuel rods. The comparison suggests that because of high axial peaking factors and relatively flat power history, fuel temperature is significantly higher in fissile zones of the RBWR-TB2 leading to higher fission gas release and cladding deformation. The higher local fuel burnup in fissile zones of RBWR-TB2 also leads to large but acceptable fuel swelling, accelerated cladding oxidation, and PCMI. In order to assist future design optimizations, sensitivity analyses on important fuel design parameters are also performed.

An evaluation of tri-valent oxide (Cr2O3) as a grain enlarging dopant for UO2 nuclear fuels fabricated under reducing environment

A study was performed to evaluate the microstructure and crystallography of nominally 500–2000 Cr2O3-doped UO2 fabricated in a temperature range of 1150–1750°C under reducing experimental conditions. An increase in grain size of the samples was observed with the increase in heat treating temperature as expected. For a given sintering temperature (1700–1750°C), an increase in the grain size was also observed with the increase in Cr2O3 concentration up to a value of ~1000–1200 wppm. A decrease in fission gas release as a function of grain size was estimated for the Cr2O3-doped UO2 samples assuming specified post-irradiation annealing conditions. A nearly linear decrease was obtained in the lattice parameter of the Cr2O3-doped UO2 fcc phase with the increase in Cr2O3 concentration, especially up to a nominal value of 1000 wppm. The lattice parameter decrease was also persistent with the increase in the average grain size as a result of addition of Cr2O3 into the UO2 lattice. An increase in the crystallite size and a decrease in the microstrain of the fcc phase were observed with the increase in the average grain size of the samples, indicating a higher crystallinity of the Cr2O3-doped samples than that of the undoped UO2 sample.

Preliminary numerical investigation of TRISO-matrix interface debonding characteristics in fully ceramic microencapsulated fuel

The bonding mode of NITE-SiC matrix and TRISO particles has a great impact on the fracture mechanisms of the Fully Ceramic Microencapsulated (FCM) fuel. Combination strength and interface debonding characteristics are of great significance when evaluating the thermo-mechanical behavior of a FCM fuel. In this study, a finite element method (FEM) model for TRISO-matrix debonding characteristics analysis was presented based on the cohesive zone model (CZM). The CZM was coupled with several irradiation behaviors (burnup, fission gas release, gap/plenum pressure, etc.) and thermo-mechanical behaviors (irradiation-induced dimensional change (IIDC), contact, heat generation, conduction, gap heat transfer, thermal expansion, and creep etc.), and developed to investigate the interface debonding characteristics of TRISO coated FCM Fuel preliminarily. A 3-dimensional multiphysics modeling approach has been successfully applied to realize simulations of a representative FCM fuel unit, also a uniformly distributed FCM layer. The NITE-SiC and OPyC interface debonding characteristics model was well performed, and was validated with CZM theory. Moreover, the effects of burnup, particle spacing, stress distributions of matrix, interfacial bonding strength and other key parameters in CZM model, were studied and investigated.

Development of a U-19Pu-10Zr fuel performance benchmark case based on the IFR-1 experiment

Metallic nuclear fuels are subject to research and development for use in advanced reactors. Robust, accurate metallic fuel performance models are important for the design, safety analysis, and licensing of these reactors. However, metallic fuel performance models require additional development; they are not as mature as UO 2 fuel performance models. A benchmark case based on the IFR-1 experiment was developed to better gauge the accuracy of existing models, identify models for high-priority development, and potentially quantify any future improvements made by further model development. This work collected publicly available information on the IFR-1 experiment and used it to develop the benchmark case. Fuel behavior during the IFR-1 irradiation was simulated by using the fuel performance code BISON, and the predicted results were compared with postirradiation examination data from the IFR-1 experiment. A sensitivity study and tuning studies were performed as a preliminary investigation into the causes of inaccurate temperature and dimensional change predictions. The benchmark predicted reasonably accurate values for the burnup and fission gas release. There was error in the predicted temperatures, which could be explained by uncertainty in the input parameters and legacy temperatures. BISON over-predicted dimensional changes in the fuel and cladding. The sensitivity study showed that the dimensional changes were most sensitive to the fuel swelling anisotropy and the cladding void swelling model. Future benchmark and model development should focus on cladding swelling behaviors to improve dimensional change predictions.

Characterization of high thermal conductivity fuel surrogates before and after ion irradiation

• Thermal conductivity of UO 2 surrogates was enhanced during sintering using aluminum oxide and silicon carbide additives. • Compositions were stable up to 5 dpa and 700 °C while a decrease in their mechanical properties was observed at 15 dpa. • Microcrack formation due to anisotropic swelling of the additives was not observed. High thermal conductivity nuclear fuels offer important potential advantages over traditional oxide-based fuels such as higher burnup, reduction in fission gas release, and better overall safety of the system. One proposed approach to high thermal conductivity fuels utilizes high thermal conductivity nonfissile additives with UO 2 fuel to lower the fuel operating temperature and thereby take advantage of the highly favorable radiation resistance of UO 2 at lower operating temperatures. However, differential swelling in the matrix and high conductivity additive phases during high dose irradiation could lead to internal cracking and poor performance. In the current study, ceria (CeO 2 ) and zirconia (ZrO 2 ) surrogate matrices were used to model UO 2 behavior. Additives of 10 vol. % Al 2 O 3 or SiC in the form of short fibers or platelets were used for the high conductivity second phase. The nuclear fuel surrogates were sintered to achieve densities greater than 93% of the ideal values. Scanning electron microscopy (SEM) imaging and X-ray diffraction confirmed the uniform distribution of the second phase and that no intermetallic second phase was formed during sintering. The thermal conductivity of the sintered samples was measured from 50 °C to 900 °C and confirmed the desirable increase compared to pure CeO 2 /ZrO 2 pellets. Samples were irradiated with 20 MeV Ni 6+ ions at midrange doses ranging from 1 to 15 displacements per atom (dpa) and temperatures from 300 °C to 700 °C. Post irradiation characterization revealed a good stability of the samples at low to medium doses with matrix lattice parameter swelling of < 0.14 % but showed a significant microstructural deterioration and decrease of the mechanical properties at 15 dpa.

Assessment of three European fuel performance codes against the SUPERFACT-1 fast reactor irradiation experiment

The design phase and safety assessment of Generation IV liquid metal-cooled fast reactors calls for the improvement of fuel pin performance codes, in particular the enhancement of their predictive capabilities towards uranium-plutonium mixed oxide fuels and stainless-steel cladding under irradiation in fast reactor environments. To this end, the current capabilities of fuel performance codes must be critically assessed against experimental data from available irradiation experiments. This work is devoted to the assessment of three European fuel performance codes, namely GERMINAL, MACROS and TRANSURANUS, against the irradiation of two fuel pins selected from the SUPERFACT-1 experimental campaign. The pins are characterized by a low enrichment (~ 2 wt.%) of minor actinides (neptunium and americium) in the fuel, and by plutonium content and cladding material in line with design choices envisaged for liquid metal-cooled Generation IV reactor fuels. The predictions of the codes are compared to several experimental measurements, allowing the identification of the current code capabilities in predicting fuel restructuring, cladding deformation, redistribution of actinides and volatile fission products. The integral assessment against experimental data is complemented by a code-to-code benchmark focused on the evolution of quantities of engineering interest over time. The benchmark analysis points out the differences in the code predictions of fuel central temperature, fuel-cladding gap width, cladding outer radius, pin internal pressure and fission gas release and suggests potential modelling development paths towards an improved description of the fuel pin behaviour in fast reactor irradiation conditions.

Three-Dimensional Modeling of Thermal-Mechanical Behavior of Accident Tolerant Fuels

Considering the safety issues of the traditional UO2-Zr fuel, a variety of accident-tolerant fuel (ATF) candidates have been proposed in recent years. Among the several ATFs, U3Si2, and UN are the two promising candidates for fuel materials owing to their high thermal conductivity and high uranium density. The FeCrAl alloy and the SiC/SiC composite material are the two promising candidates for cladding owing to their high oxidation resistance and high strength. In order to quantitatively evaluate the performance of ATFs, this study summarizes the physical models of typical ATF cladding materials (FeCrAl and SiC) and pellet materials (UN and U3Si2). Then a three-dimensional non-linear finite element method is applied to simulate the thermal-mechanical behavior of several typical fuel-cladding combinations, including UO2-FeCrAl, UN-FeCrAl, U3Si2-FeCrAl, U3Si2-Zr, and U3Si2-SiC. The important physical quantities, such as the fuel centerline temperature, the deformation of the pellet and the cladding as well as the pellet-cladding mechanical interaction (PCMI) were studied. The fission gas release model was also verified and improved.