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Under transient conditions, rapid fuel swelling in UO2 can occur through fission gas and vacancy capture by intra-granular bubbles. These swellings can lead to significant cladding deformations and clad failure. To understand the mechanisms of intra-granular swelling a number of key experiments have been conducted by the UK in conjunction with the Halden Reactor. Post-irradiation annealing enables the study of the thermal behaviour of the intra-granular bubbles in the absence of complicating processes such as irradiation-induced bubble destruction and fission gas resolution. Bubble-like features are frequently observed in straight lines and it has always been assumed that these were nucleated in the wake of energetic fission fragments and served as nuclei for bubble growth. However, most of these bubbles do not grow significantly and viable intra-granular bubbles are nucleated at much lower densities. At the temperature and pressure that these form, the xenon and krypton are above their critical points and the elements are supercritical fluids. However, bubble growth occurs at temperatures and pressures much greater than the critical points and the gases may be treated with simple equations of state such as those by Clausius or Van der Waals. Of greater importance is the fact that the large Van der Waals forces in these atoms generated by fluctuations in the electron dispositions lead to extensive adsorption on the inner surfaces of the bubbles and vacancy capture by the bubbles can lead to the diffusive loss of fission gas atoms resulting in bubble growth at slight over-pressure conditions. A diffusion model based on Langmuir Adsorption has been developed and shown to be consistent with the out-of-pile annealing studies. [ABSTRACT FROM AUTHOR]

Under transient conditions, rapid fuel swelling in UO2 can occur through fission gas and vacancy capture by intra-granular bubbles. These swellings can lead to significant cladding deformations and clad failure. To understand the mechanisms of intra-granular swelling a number of key experiments have been conducted by the UK in conjunction with the Halden Reactor. Post-irradiation annealing enables the study of the thermal behaviour of the intra-granular bubbles in the absence of complicating processes such as irradiation-induced bubble destruction and fission gas resolution. Bubble-like features are frequently observed in straight lines and it has always been assumed that these were nucleated in the wake of energetic fission fragments and served as nuclei for bubble growth. However, most of these bubbles do not grow significantly and viable intra-granular bubbles are nucleated at much lower densities. At the temperature and pressure that these form, the xenon and krypton are above their critical points and the elements are supercritical fluids. However, bubble growth occurs at temperatures and pressures much greater than the critical points and the gases may be treated with simple equations of state such as those by Clausius or Van der Waals. Of greater importance is the fact that the large Van der Waals forces in these atoms generated by fluctuations in the electron dispositions lead to extensive adsorption on the inner surfaces of the bubbles and vacancy capture by the bubbles can lead to the diffusive loss of fission gas atoms resulting in bubble growth at slight over-pressure conditions. A diffusion model based on Langmuir Adsorption has been developed and shown to be consistent with the out-of-pile annealing studies. [ABSTRACT FROM AUTHOR]

The UMo-Al monolithic fuel element has a special failure mode of blistering on the cladding surface under high temperature and high burn-up conditions. In order to provide an auxiliary means for the safety analysis and failure limit formulation of the fuel element, this study establishes a numerical simulation method of fuel failure behavior, considering the fracture behavior and the fission gas pressure acting on the cracks for the UMo-Al monolithic fuel element. Numerical simulation is based on ABAQUS software and the extended finite element method (XFEM) established by programming FORTRAN subroutines UMAT, UMATHT, and UAMP and using the thermo-mechanical sequential coupling method. Combined with Python to secondary develop ABAQUS with the FORTRAN subroutine, this study realizes the coupling of fission gas pressure load and crack propagation. This study obtained the failure threshold temperature, blistering height, area, and other blistering characteristics of the UMo-Al monolithic fuel element through numerical simulation of annealing experiments on L1P460 in RERTR-12. A comparison of the numerical simulation results with the RERTR-12 experiment shows that the method established in this study can effectively analyze blistering failures. [ABSTRACT FROM AUTHOR]

Large-grain UO 2 has attracted much attention due to its advantage of suppressing fission gas release by reducing the specific surface area of grain boundaries. In this study, a single grain growth additive MnO is selected to modify UO 2 by conventional pressureless sintering method. The effect of MnO modification on UO 2 densification and grain growth at temperatures between 1200 and 1700 °C was investigated in detail. 0.5 wt%MnO-UO 2 pellets can reach nearly 95% T.D. when sintered at 1200 °C for 24 h. An average grain size greater than 300 µm can be obtained at 1700 °C for 24 h. A series of sintering experiments were designed and conducted to verify the specific mechanism of MnO on the grain growth of UO 2. The classical grain growth kinetics indicates that the grain growth exponent of MnO modified UO 2 in the early stage is 3, which corresponds to the lattice diffusion under pore control, while that of UO 2 is 2 and conforms to the grain boundary migration control. The SEM results show that there is solute segregation of MnO at the grain boundaries, followed by nucleation of lamellar structure in the form of piles and plates on the boundaries after high temperature treatment and traces of liquid phase in some region. Diffusion sintering experiments show that the mechanism of MnO-promoted UO 2 grain boundary migration is based on solid phase diffusion. The XRD results of UO 2 modified with different amounts of MnO demonstrate that the growth degree of UO 2 grains can be related to the variation of lattice parameters resulted from the MnO modification. This is due to the enhanced mobility of UO 2 grain boundaries because of the large distortion of surrounding lattice caused by Mn modification. These results can provide new ideas for the manufacture of UO 2 ceramic fuels. [ABSTRACT FROM AUTHOR]

Focussed ion beam - scanning electron microscope (FIB-SEM) 3D examination was conducted on three standard UO2 and one Cr doped UO2 high burn-up pressurized water reactor (PWR) fuel samples. This work complemented other microanalysis examination, including an electron backscattered diffraction (EBSD) work on the polished surface. A parallel article giving the EBSD results was submitted simultaneously. Together, they found, in all the central area of these high burn-up samples: (i) a restructuring of the initial grains into smaller sub-grains forming low angle boundaries and with crystal orientations around that of their parent grains; and (ii) intragranular bubbles mostly situated on these low angle boundaries. The FIB-SEM 3D examination showed how such inter-sub-grain bubbles start as small compact but also small lenticular bubbles, similar to typical small intergranular lenticular bubbles. With increasing burn-up, these lenticular bubbles get thicker and locally interlink to form more complex bubbles. However, no long distance networks, between the sub-grains or between the original grains, were found. Such networks could have been a path for part of the fission gases to reach the grain boundaries, the grain edges (the intersection line of three grain boundaries), and the rod free volumes. These FIB-SEM 3D examinations brought details on the intragranular and intergranular bubbles situation for each studied volume. The distribution of the intragranular bubbles according to their sizes and shapes was exposed. The central restructuring, studied in this work, is likely to play a role in the increase of the fission gas release fractions at high burn-up. This work is an incentive to study further this restructuring and the bubbles formed, combining different approaches. [ABSTRACT FROM AUTHOR]

Advanced reactor technologies are being considered for the next-generation of nuclear power plants. These plants are designed to have a smaller footprint, run more efficiently at higher temperatures, have the flexibility to meet specific power or heating needs, and have lower construction costs. This paper offers a perspective on molten salt reactors, promoted as having a flexible fuel cycle and close-to-ambient pressure operation. A complexity introduced by reducing the reactor footprint is that it may require low-enriched fuel for efficient operation, available from enrichment of the feed salt or by reusing actinides from existing used nuclear fuel (UNF). Recycling UNF has the potential to reduce high-level waste, if done correctly. Release limits from UNF processing are stringent, and processes for waste reduction, fission gas trapping, and stable waste-form generation are not yet ready for commercial deployment. These complex processes are expensive to develop and troubleshoot because the feed is highly radioactive. Thus, fuel production and supply chain development must keep abreast of reactor technology development. Another aspect of reactor sustainability is the non-fuel waste streams that will be generated during operation and decommissioning. Some molten salt reactor designs are projected to have much shorter operational lifetimes than light-water reactors: less than a decade. A goal of the reactor sustainability effort is to divert these materials from a high-level waste repository. However, processing of reactor components should only be undertaken if it reduces waste. Economic and environmental aspects of sustainability are also important, but are not included in this perspective. [ABSTRACT FROM AUTHOR]

Identification of a failed fuel subassembly due to the dry rupture of its fuel pin in prototype fast breeder reactor is one of the safety protocols to be standardized for its safe operation for which interaction of fission gases with liquid sodium was studied. This paper reports the studies on the feasibility of detection of fission gases in presence of sodium as well as separation of fission gases out of liquid sodium through Ar purge in a prototypical sodium module with 300 mL of Na. The resultant gases were adsorbed on activated charcoal and determined the activity of fission gases using high resolution gamma spectrometry. Quantitative recovery of fission gases from sodium matrix in the range of sub-picogram levels was established. [ABSTRACT FROM AUTHOR]

Diffusion in bcc uranium and U–Mo alloys is of great interest because fission gas and other fission products impact the performance of nuclear fuels. We investigate the mobility of xenon and molybdenum in bcc uranium (γ -U) and metallic U–Mo alloys by calculating the migration energies of xenon and molybdenum for various local compositions using density functional theory. We also calculate the solute–vacancy binding energies of different solutes to vacancies in bcc uranium. We find that the solute–vacancy binding energy in bcc uranium is significantly higher than it is in other bcc metals (e.g., Fe and W). We also find that the migration energy of molybdenum is substantially higher than the migration energy of xenon, indicating that xenon is much more mobile than molybdenum in bcc uranium. The presence of molybdenum in the nearest-neighbor shell around a xenon atom typically increases the migration energy of xenon, which indicates a reduction of xenon mobility in U–Mo alloys compared to pure bcc uranium. [ABSTRACT FROM AUTHOR]

U-10 wt.% Zr (U-10Zr) metallic fuel is the leading candidate for next-generation sodium-cooled fast reactors. Porosity is one of the most important factors that impacts the performance of U-10Zr metallic fuel. The pores generated by the fission gas accumulation can lead to changes in thermal conductivity, fuel swelling, Fuel-Cladding Chemical Interaction (FCCI) and Fuel-Cladding Mechanical Interaction (FCMI). Therefore, it is crucial to accurately segment and analyze porosity to understand the U-10Zr fuel system to design future fast reactors. To address the above issues, we introduce a workflow to process and analyze multi-source Scanning Electron Microscope (SEM) image data. Moreover, an encoder-decoder-based, deep fully convolutional network is proposed to segment pores accurately by integrating the residual unit and the densely-connected units. Two SEM 250 × field of view image datasets with different formats are utilized to evaluate the new proposed model's performance. Sufficient comparison results demonstrate that our method quantitatively outperforms two popular deep fully convolutional networks. Furthermore, we conducted experiments on the third SEM 2500 × field of view image dataset, and the transfer learning results show the potential capability to transfer the knowledge from low-magnification images to high-magnification images. Finally, we use a pre-trained network to predict the pores of SEM images in the whole cross-sectional image and obtain quantitative porosity analysis. Our findings will guide the SEM microscopy data collection efficiently, provide a mechanistic understanding of the U-10Zr fuel system and bridge the gap between advanced characterization to fuel system design. [ABSTRACT FROM AUTHOR]

Gaseous fission products from nuclear fission reactions tend to form fission gas bubbles of various shapes and sizes inside nuclear fuel. The behavior of fission gas bubbles dictates nuclear fuel performances, such as fission gas release, grain growth, swelling, and fuel cladding mechanical interaction. Although mechanical understanding of the overall evolution behavior of fission gas bubbles is well known, lacking the quantitative data and high-level correlation between burnup/temperature and microstructure evolution blocks the development of predictive models and reduces the possibility of accelerating the qualification for new fuel forms. Historical characterization of fission gas bubbles in irradiated nuclear fuel relied on a simple threshold method working on low-resolution optical microscopy images. Advanced characterization of fission gas bubbles using scanning electron microscopic images reveals unprecedented details and extensive morphological data, which strains the effectiveness of conventional methods. This paper proposes a hybrid framework, based on digital image processing and deep learning models, to efficiently detect and classify fission gas bubbles from scanning electron microscopic images. The developed bubble annotation tool used a multitask deep learning network that integrates U-Net and ResNet to accomplish instance-level bubble segmentation. With limited annotated data, the model achieves a recall ratio of more than 90%, a leap forward compared to the threshold method. The model has the capability to identify fission gas bubbles with and without lanthanides to better understand the movement of lanthanide fission products and fuel cladding chemical interaction. Lastly, the deep learning model is versatile and applicable to the micro-structure segmentation of similar materials. [ABSTRACT FROM AUTHOR]

The present paper considers the processes occurring in a failed fuel element during fuel cladding failure detection in the mast sipping system of the refueling machine. The time taken for water and fission gas products to leak from a failed fuel element during its lifting was estimated depending on the size of the defect. A period of 10–12 min exposure prior to bubbling was demonstrated to be applicable for detecting defects with a size of about 30 μm. The sizes of gas bubbles leaving the fuel element during the failure detection were estimated along with their rate of ascent. An algorithm for conducting fuel element failure detection that increases the sensitivity of the method without extending the refueling is proposed. [ABSTRACT FROM AUTHOR]

He-Xe gas-cooled micro-molten salt reactor (HX-micro-MSR) is a novel reactor concept that integrates the technologies of molten salt reactor and closed Brayton cycle (CBC), which has a more compact structure compared with the other types of micro-MSRs. The reactor system probed in this paper has a thermal power of 2.5 MWth and consists of the core, fission gas collecting container, and CBC system. The He-Xe gas flows through the coolant channels in the core to transfer the nuclear heat to the CBC for electricity production. To withstand the large pressure difference between the He-Xe gas coolant (7 MPa) and molten salt (nearly atmospheric pressure), the sizes and wall thickness of coolant channels are analyzed and optimized to be 8 mm and 0.5 mm, respectively. In total, 1050 coolant channels are regularly arranged in the core to effectively remove the nuclear heat. The enrichment of U-235 in the fuel UF4 is determined to be 61% to meet the burnup goal of 5 years. To minimize the neutron leakage, the thickness of reflector is selected to be 23 cm. Six control drums with the diameter in 10.5 cm and neutron absorber layer thickness in 1 cm are symmetrically placed in the reflector to realize reactivity control under both normal and accident conditions. Radiation shield is designed and analyzed by varying the thickness of multiple shielding layers, and it is found that a combination of W layer in 14 cm and Mg(BH4)2 layer in 56 cm can satisfy the criterion of radiation dose (<1 mSv/h). The total weight of HX-micro-MSR is estimated to be 35 tonnes and falls into the acceptable range. [ABSTRACT FROM AUTHOR]

Fission gas plays a significant role in fuel rod performance following accidents. The amount of fission gas increases dramatically under accidental conditions. This leads to a subsequent rise in the fuel rod internal pressure and temperature due to aggravated gap conductance between the fuel pellet and cladding. As a result, fuel rod performance degrades. Therefore, studying fission gas behavior is crucial for accident assessment and evaluating fuel rod performance. Minimizing the Impact of fission gas on fuel rods is essential for maintaining their integrity and safety within nuclear reactors. One important aspect of ensuring safety is predicting fission gas release (FGR). In this study, we presented an extended model to be used in light water reactors (LWRs). The FGR can be modeled using COMSOL Multiphysics with the finite element method. This modeling approach considers both normal and abnormal conditions, with the latter categorized as Class-II type incidents. The model assumes that the gas diffusion inside a spherical grain varies over time. By examining perfect sinks with gas production, perfect sinks without gas production, and imperfect sinks under steady-state conditions, different initial and boundary conditions are set. To validate the accuracy and universality of expressions used in the model, input parameters from other models and experiments are utilized. By comparing the model's results with these inputs, the accuracy and applicability of the expressions can be confirmed. This validation process ensures that the model provides reliable predictions for fission gas behavior in fuel rods under both normal and abnormal operating conditions. Based on our findings, it is evident that the FGR fraction displays an upward trend as diffusion coefficients and temperatures rise. Conversely, larger grain sizes and higher linear heat generation rates are associated with a reduction in the FGR fraction. Notably, enhanced resolution leads to a postponed onset of FGR. Furthermore, the influence of the diffusion coefficient on the FGR fraction primarily stems from the interconnected effects of temperature and linear heat generation rate. [ABSTRACT FROM AUTHOR]

High fidelity simulation of fission gas behavior is able to help us understand and predict the performance of nuclear fuel under different irradiation conditions. Cluster dynamics (CD) is a mesoscale simulation method which is rapidly developed in nuclear fuel research area in recent years, and it can effectively describe the microdynamic behavior of fission gas in nuclear fuel; however, due to the huge cost of computation needed for CD model solution, the application scenario of CD has been limited. Thus, how to design the acceleration algorithm for the given computing resources to improve the computing efficiency and simulation scale has become a key problem of CD simulation. In this work, we present an accelerating cluster dynamics model based on the spatially dependent cluster dynamics model, combined with multi optimization methods on a DCU (deep computing unit)-based heterogeneous architecture supercomputer. The correctness of the model is verified by comparing with experimental data and Xolotl—a software of SciDAC program from the U.S. Department of Energy's Office of Science. Furthermore, our model implementation has a better computing performance than Xolotl's GPU version. Our code has gained great strong/weak scaling performance with more than 72.75%/84.07% parallel efficiency on 1024 compute nodes. This work developed a new efficient model for CD simulation of fission gas in nuclear fuel. [ABSTRACT FROM AUTHOR]

Fission gas isotopic compositions are sensitive to a variety of reactor operating parameters that include the neutron flux, neutron fluence at different neutron energies, and operating temperature. Measurements of fission gas isotopic compositions thus have potential to constrain reactor simulations for nuclear forensics and safeguards applications. In this paper, we present Kr and Xe isotope measurements from a suite of samples obtained from locations that span the axial length a fuel pin with a well-characterized irradiation history and compare these data to spatially resolved reactor simulations. We observed positive correlations between fluence sensitive isotopic ratios and burnup and between a flux sensitive ratio and power, although some discrepancies are observed between the measured data and model predictions. These differences may be due to simplifications in the model and/or inaccuracies in the cross sections. A much broader measurement to model comparison is required to better understand the discrepancies. [ABSTRACT FROM AUTHOR]

We present a generalised approach for describing the universal dynamics and statistical properties of fission gas bubble growth in amorphous systems, using U 3 S i 2 as an example. We examine the scaling properties of both the growth and coalescence regimes of bubble growth, and find that the scaling exponents in these regimes are controlled by the fission rate and temperature. The nucleation regime is characterised by a growth exponent close to 1/2, while the coalescence regime is characterised by a growth exponent around 1/3. We show that the statistical distribution over bubble size is of a universal character. Analytical results are verified by numerical simulations and clarified with experimental data. This study provides insight into the kinetics of fission gas bubble growth in amorphous systems. [ABSTRACT FROM AUTHOR]

A ceramic–ceramic (CERCER) fuel with minor actinide-enriched ceramic fuel particles dispersed in a MgO ceramic matrix is chosen as a promising composite target for accelerator-driven systems (ADS). Fission swelling is a complex irradiation-induced phenomenon that involves recrystallization, resolution, and hydrostatic pressure under extreme conditions of high temperature and significant fission flux. In this study, a multiscale computational framework was developed to integrate simulations of continuum-scale thermo-mechanical behavior in the CERCER composite with a grain-scale hydrostatic pressure-dependent fission gas swelling model. Hydrostatic pressure-dependent fission welling is taken into account in the stress update algorithms for U O 2 particles. Accordingly, we programmed the user subroutines to define the thermo-mechanical constitutive relations in the finite element simulations. The obtained results indicate that (1) the proposed method accurately predicts the swelling deformation at various burnup levels while taking into account hydrostatic pressure and (2) prior to recrystallization, the particle swelling is primarily influenced by temperature variation, whereas after recrystallization, the presence of hydrostatic pressure favorably suppresses the swelling deformation. This work effectively captures the swelling behavior influenced by hydrostatic pressure within the dispersed-type CERCER composite fuel in ADSs. [ABSTRACT FROM AUTHOR]

Taking inventories in reactor cores is critical for understanding their radioactive source terms and establishing the relationship between the activity concentration in the primary loop and the status of the reactor core's fuel. However, there is a niche in which a simple but accurate relationship between reactor conditions and nuclide inventories can reliably predict the fission gas nuclide activities of the reactor core in the primary loop. In this study, a simple and efficient model called "Inventories of a Point Reactor for Fission Gas Nuclides" (IPRFGN) was proposed to calculate and interpret such inventories, in which a 10 MW high-temperature gas-cooled experimental reactor (HTR-10) was used as the test case. The present study findings were consistent with those of a general point–depletion burnup code such as the KORIGEN code. Here, the relative error was <1%. Based on the application of the IPRFGN model in HTR-10, the results indicate that the proposed IPRFGN model has provided the relationship between the inventories of fission gas nuclides in the core and the reactor conditions in all types of nuclear fission reactors. In the future, the IPRFGN model will be used for calculating fission gas nuclide inventories in various reactors. [ABSTRACT FROM AUTHOR]

Molten salt reactor operation will necessitate circulation of a cover gas to remove certain evolved fission products and maintain an inert atmosphere. The cover gas leaving the reactor core is expected to contain both noble and non-noble gases, aerosols, volatile species, tritium, and radionuclides and their daughters. To remove these radioactive gases, it is necessary to develop a robust off-gas system, along with novel sensors to monitor the gas stream and the treatment system performance. In this study, a metal organic framework (MOF) was engineered for the capture of Xe, a major contributor to the off-gas source term. The engineered MOF column was tested with a laser-induced breakdown spectroscopy (LIBS) sensor for noble gas monitoring. The LIBS sensor was used to monitor breakthrough tests with various Xe, Kr, and Ar mixtures to determine the Xe selectivity of the MOF column. This study offers an initial demonstration of the feasibility of monitoring off-gas treatment systems using a LIBS sensor to aid in the development of new capture systems for molten salt reactors. [ABSTRACT FROM AUTHOR]

The expected benefits from Cr-doped fuel are improved retention of fission gases within the pellets due to its large grain size. To demonstrate this, several experiments have been carried out by Halden reactor and Studsvik. These experiments are now being used to benchmark several fuel performance codes among them transuranus code. All this as part of a Coordinate Research Project (CRP) by IAEA named Testing and Simulation for Advanced Technology and Accident Tolerant Fuels (ATF-TS). This work is introducing a novel fission gas diffusivity model for doped fuel in transuranus code. It is observed the benefits of introducing this new model when comparing to the standard model already existing in transuranus. Nevertheless, more work needs to be carried out to fully understand all the phenomena involved in adding dopant in UO2 due to change of thermo mechanical properties. [ABSTRACT FROM AUTHOR]

An integrated effort by the Versatile Test Reactor (VTR) Gas-Cooled Fast Reactor (GFR) Team to develop an experiment vehicle or extended-length test assembly for the VTR experiments is led by Idaho National Laboratory and supported by an industrial partner, General Atomics, and university partners, including Texas A&M University, University of Michigan, Oregon State University, University of Houston, and University of Idaho. The focus of the effort is to design a helium gas-cooled cartridge loop (GCL) to assist with the testing of fuels, materials, and instrumentation to further support development of advanced reactor systems. This study is divided into two parts. Part I provides the functional requirements and critical irradiation data needs for advancing gas-cooled fast reactor (GFR) technologies. The objective of Part I is to describe the overall preliminary conceptual design of the VTR helium cartridge loop, the design of a fission product venting system, thermal-hydraulic effects of flow direction, and gamma-heating generation in the cartridge. This paper, Part II, includes the measurement techniques being developed to measure the thermophysical properties of the different materials that make up the GCL, as well as the instrumentation and control system within the cartridge required for advancing GFR technologies. The purpose of Part II is to describe the functionality and efficacy of the measuring systems being developed to support the GCL. These systems include (a) a unique measurement platform that joins ion irradiation and a laser beam with an infrared camera and X-ray detection equipment developed and used to investigate more accurately and efficiently the influence of radiation and fission gases on the material properties under high temperatures; (b) a laser-induced breakdown spectroscopy to demonstrate its capability of monitoring possible fuel failure by detecting sub–part-per-million levels of xenon in the helium coolant stream, providing experimental data to better understand the interactions of fuel elements and coolant at high temperature, pressure, and fast neutron flux; (c) fiber-optic sensors with the ability to measure both the temperature demonstrated using a three-dimensional printed heat exchanger and, potentially, the strain in harsh environments; and (d) surface emissivity measurement test rigs to understand the effect of temperature, radiation, and surface finish on the silicon carbide cladding surface emissivity. Additional analyses and development, as well as integrated out-of-pile testing, are planned to demonstrate and validate the accuracy of the measuring systems and instrumentation in a more prototypic environment prior to their implementation into the VTR. [ABSTRACT FROM AUTHOR]

Metallic fuels, particularly U 10Zr, are promising candidates for next-generation sodium-cooled fast reactors. Irradiation of nuclear fuels in reactors can lead to the formation of solid and gas fission product which subsequently forms microstructural pores, deteriorating fuel performance. Due to the massive amount of pores and complex phases formed, a quantitative description of fission gas pores is not yet available, preventing the development of microstructure-informed fuel performance modeling for fuel qualification. This paper applied a pre-trained deep learning model to ∼10,260 high magnification scanning electron microscopy images. This method increased the accuracy of fission gas pore segmentation and allows statistical features to be extracted which cannot be achieved manually. A pre-trained decision tree model worked on the segemenation resutls and further classified the pores into different categories to produce a correlation between the pores, movement of lanthanides, and temperature gradient during irradiation. This paper emphasizes the potentials of machine learning models to accelerate fuel research, development, and qualification for advanced reactors. [Display omitted] • Deep learning model is used to detect pores in ∼10,260 scanning electron microscopy images. • About 230,000 pores are detected and classified based on their impact on nuclear fuel performance. • Pore statistics and orientations help to understand fuel performance from post-irradiation examination. [ABSTRACT FROM AUTHOR]

We present the results of a first assessment of the multiphysics PLEIADES/GERMINAL V3 fuel performance code against integral irradiation experiments of americium-bearing MOX fuel pins irradiated in sodium fast reactors. The pins considered in this work cover a range of americium contents from 2 to 5 weight %, being representative for the homogeneous transmutation path. The code predictions are compared to several post-irradiation examination results on integral (e.g., fission gas release) and local quantities (e.g., actinides redistribution in the fuel pellet), and demonstrate a satisfactory agreement with experimental data. The deviations between calculated and experimental quantities shed some light on modeling developments needed to better calculate the behavior of americium-bearing mixed-oxides fuel pins under irradiation. • Assessment of the PLEIADES/GERMINAL V3 code against transmutation irradiation experiments. • The experiments considered encompass low and high burnup fuel pins. • Code predictions are compared to fuel pin integral and local experimental data. • The comparisons show a good predictive capability on quantities such as fission gas release and central hole formation. • Needed modeling advancements are identified, especially on the radial transport of acitinides and axial cladding deformation. [ABSTRACT FROM AUTHOR]

Accelerating the nuclear fuel qualification process will rely on some combination of advanced modeling and simulation techniques with accelerated irradiation testing and separate effects experiments to enable the development of new fuel concepts in a shorter time frame. One of the key challenges to successfully leveraging accelerated irradiation tests will be understanding the artifacts that may be introduced with accelerated accumulation of dose and/or burnup. This work presents phase field (MARMOT) simulations of the evolution of representative 2D UO 2 microstructures up to 40 MWd/kgU. Simulations were performed under both commercial light water reactor fuel conditions as well as those that would be expected for highly accelerated (∼10x) burnup conditions similar to those used in the MiniFuel irradiations in Oak Ridge National Laboratory's High Flux Isotope Reactor. The phase field model was coupled with a discrete nucleation algorithm to model restructuring at high burnup. The effect of the different fission rates in both microstructures was investigated at two temperatures: 650 ∘C and 800 ∘C. The lower temperature simulations both showed an onset of restructuring at nearly 60 MWd/kgU. More extensive restructuring was obtained in the MiniFuel microstructure compared with that of the PWR fuel. At 800 ∘C, no restructuring was obtained as a result of the thermally activated diffusion of Xe atoms and U vacancies to fission gas bubbles, which reduces the nucleation driving force. These results highlight the importance of using modeling and simulation tools to inform the environmental conditions during targeted accelerated irradiation tests to extract the most useful fuel performance data. [ABSTRACT FROM AUTHOR]

Moyens d’Etude des Rejets de Gaz de fission par Recuit (MERARG) experimental facility in LECA-STAR Laboratory at the CEA Cadarache center allows characterizing nuclear fuels with respect to the behavior of fission gases during thermal transients. In MERARG, fission gas release (F.G.R.) measurement is carried out remotely from the sample where the release occurs, in a dedicated counting chamber located outside the hot cell itself. It is therefore necessary to deconvolve the acquired experimental data in order to infer the release kinetics at the sample position. An analytical deconvolution method allows us to determine, from the experimental data, the kinetics of the F.G.R. at the sample level located in the MERARG furnace with, in some few cases, a margin of error. It is established that the F.G.R. takes the form of successive bursts depending, in part, upon temperature and characteristics of the fuel sample. For the better understanding of the mechanisms at the origin of the characteristic kinetics of F.G.R., this article aims to expose a new mathematical method for reprocessing the experimental data obtained using a forward-fitting approach, allowing us to overcome the difficulties encountered with the analytical method. This forward-fitting approach poses the question of the nature of the mechanisms underlying F.G.R. as an inverse problem. By making assumptions about these mechanisms, the forward-fitting method allows the F.G.R. interpretation by comparing them to analytically deconvoluted F.G.R. signals. This method provides a different way to analyze the annealing test data, making it possible to describe physical phenomena directly related to the gas release from the samples, such as the sudden release occurring at different temperatures. [ABSTRACT FROM AUTHOR]

The thermal conductivity of U-Mo alloys for research reactor fuel was studied. Thermal conductivity is one of the most important properties concerning design, performance analysis, and safety evaluation of these alloys as research reactor fuel. Thermal conductivity data of unirradiated U-Mo alloy collected from literature were examined, analyzed, and used to develop a correlation as a function of Mo content and temperature. For irradiated U-10Mo alloy, a thermal conductivity correlation was developed as a function of fission density and temperature. This model considers the effects of porosity growth by fission gas bubbles, solid fission product accumulation, and buildup of grain boundaries due to recrystallization during irradiation. [ABSTRACT FROM AUTHOR]

This work follows the AGR-like FHR ongoing research that proposes to adopt the British Advanced Gas cooled Reactor (AGR) geometry combined with the molten salt Fluoride-Cooled High Temperature Reactor (FHR) concept. This work presents the new models and material properties implemented in the TRANSURANUS code for Cr-doped UO 2 fuel, which is one of the advanced technology fuels considered for light water reactors. For this purpose, we update the mechanistic model for fission gas behaviour in the code by means of a dedicated fission gas diffusion coefficient recently proposed by Cooper et al. on the basis of atomistic scale simulations to take into consideration the impact of the dopant on the point defects that control the fission gas diffusivity in various temperature regions of interest. In a consistent manner, we propose also a modified creep correlation based on the mechanistic model for standard oxide fuels. Furthermore, we analyse the effect of cracking observed in doped fuels subjected to power ramps and take into consideration the limited densification of the high-density fuel reported in the open literature. A subsequent parametric study pointed out the main factors affecting the integral fission gas release rate, which was shown to be a limiting factor in the AGR-like FHR under consideration. Finally, the improved performance of the advanced technology fuel is shown by means of the reduced inner gas pressure at end of life, as well as the reduced pellet cladding mechanical interaction during the postulated operational transient from the open literature for AGRs. As a result, the doped fuel is shown to be able to sustain higher power levels in the AGR-like FHR. • Introduction of Cr-doped properties into TRANSURANUS: diffusion coefficient, creep and cracking. • Assessment of the effect of fission gas behaviour models. • Parametric study of Cr-doped correlations in TRANSURANUS. • Assessment of the effect of Cr-doped fuel in the AGR-like FHR during normal and off-normal operation. [ABSTRACT FROM AUTHOR]

Fuel performance codes are used to forecast fuel behavior and ensure safe operation. These analyses must typically include prediction uncertainties, and fuel performance models need calibration. Consequently, code calibration must derive the best estimates and corresponding uncertainties of model parameters for subsequent propagation. Bayesian calibration is popular for generating the probability distribution of model parameters. However, model inadequacy disrupts these techniques, typically resulting in underestimated uncertainties. Earlier research showcased the incorporation of model inadequacy by model parameter inflation. The method demands cheap code predictions and derivatives, which required further research to develop differentiated Gaussian process surrogates. This work combines those techniques into a complete methodology. We demonstrate it by calibrating Transuranus against fission gas release and cladding oxidation data. The result is model parameter uncertainties that primarily explain the discrepancies between the predictions and corresponding measurements, except when the output behaves highly non-linearly. • Calibration of Transuranus with fission gas release and cladding oxidation data. • Model inadequacy disrupts calibration, leading to underestimation of uncertainties. • Derivative-based model parameter uncertainty inflation accounts for model inadequacy. • Gaussian Process surrogates aid calibration with cheap predictions and derivatives. • The methodology works in linear cases but falters in heavy non-linear applications. [ABSTRACT FROM AUTHOR]

Uranium mononitride (UN) is one of the ceramic nuclear fuel alternatives to oxide fuel considered for light water reactors and advanced reactor designs. Properties like self- and fission gas diffusivity need to be better understood, given that they influence key fuel performance phenomena such as fission gas swelling and release. In particular, the radiation induced athermal (D 3) diffusivity remains challenging to accurately predict and has only been sparsely characterized in UN, despite its importance as it likely governs diffusion at the low temperatures this high-thermal-conductivity fuel form may operate. Molecular Dynamics simulations are used to estimate the mean square displacement induced by a primary knock-on atom (PKA) with a given kinetic energy. These results are combined with the PKA energy distributions obtained from binary collision approximation calculations to obtain the displacement due to a particular fission fragment. Finally, this is combined with experimental fission fragment yields to determine the displacement due to an average fission event and, thus, express the athermal diffusivity as a function of the fission rate density. These results are in excellent agreement with available experimental data. A particular importance is given to the understanding and the quantification of the variability of these results. [ABSTRACT FROM AUTHOR]

Increasing the average grain size of fuel pellets by doping them with chromium oxide is one strategy to improve oxide nuclear fuels performance. The promoted fission gas retention is thought to improve the performance of the fuel at high burnup. In this work, we review models for the solubility of chromium in UO 2 , and the evolution of the chromium phases in the fuel matrix during irradiation. These models are implemented in SCIANTIX, an open-source mesoscale code describing inert gas behaviour in nuclear fuel. We adjusted the chromium solubility model keeping each parameter within its range of compatibility with experimental data, targeting a better representation of available electron probe microanalysis data of chromium content in fuel after irradiation. As for fission gas behaviour, we considered a physics-based description of the chromium impact on the fission gas diffusivity in fuel grains. The expression for the fission gas diffusivity in standard non-doped uranium oxide has been extended by introducing the impact of the concentration of defects introduced by interstitial oxygen excess representing the effect of chromium content in the fuel itself. A preliminary integral assessment of the proposed models has been carried out against the available experimental data. • Chromium-doped oxide fuels target increased safety and performance. • Chromium-doping increases fuel grain size and promotes effective retention of gases. • Recent experimental findings highlighted the phases arising in chromium-doped fuels. • An engineering model linking phase evolution and fission gas diffusivity is proposed. • The proposed model is tested in SCIANTIX and targets application in engineering tools. [ABSTRACT FROM AUTHOR]

Mesoscale simulations of fission gas release (FGR) in nuclear fuel provide a powerful tool for understanding how microstructure evolution impacts FGR, but they are computationally intensive. In this study, we present an alternate, data-driven approach, using deep learning to predict instantaneous FGR flux from 2D simulated nuclear fuel microstructure images. Four convolutional neural network (CNN) architectures with multiscale regression are trained and evaluated on simulated FGR data generated using a hybrid phase field/cluster dynamics model. All four networks show high predictive power, with R 2 values above 98%. The best performing network combines a Convolutional Block Attention Module (CBAM) and InceptionNet mechanisms to provide superior accuracy (mean absolute percentage error of 4.4%), training stability, and robustness on very low instantaneous FGR flux values. The trained model demonstrates the utility of deep learning in this context, but has no validity outside the 2D model and conditions used to generate the training data. [ABSTRACT FROM AUTHOR]

Phase-field modeling has increasingly been used to study microstructural evolution in fission gas bubbles in nuclear fuel to improve understanding of fission gas release. To improve computational efficiency, often only vacancies and gas atoms are included as defect species. In this case, the net effects of vacancy and interstitial production, recombination, and biased sink absorption are included as a net vacancy source, or net vacancy source combined with an effective sink. However, there has been a lack of clarity on what parameter values should be used for these approaches to best match the more complete physical picture that includes interstitials and vacancies. Here, we compare a phase-field model of void growth to analytical models for the source-only and source plus sink approach to gain insight into how the phase-field models can be parameterized effectively. The source-only approach provides greater flexibility to match growth rates determined from the full vacancy-interstitial picture. A strategy was developed for determining the value of the net vacancy source term by comparing to an analytical model that includes vacancy and interstitial production, recombination, and biased sink absorption. This strategy can be used to parameterize phase-field models of fission gas bubble growth. • Net vacancy production can be approximated by vacancy source-only and source+sink approaches. • Quasi-steady state analytical models explain behavior of phase-field models. • Vacancy source-only models have more flexibility to match vacancy-interstitial models. • Fitting source term in source-only model provides a good approximation to full vacancy-interstitial model. [ABSTRACT FROM AUTHOR]

Mechanistic multi-scale modelling holds the potential to inform fuel performance codes by incorporating high-fidelity models, algorithms, parameters, and material properties. In this context, meso-scale codes emerge as valuable tools for developing detailed models and performing separate verification and validation steps. This work focuses on SCIANTIX, an open-source 0D meso-scale code designed to describe the behaviour of gaseous and volatile fission products in nuclear oxide fuel. The code predominantly employs engineering physics-based behavioural models featuring computational times that align with typical fuel performance code requirements. Given the numerical foundation of the code, it is applicable to both stationary and transient conditions. Following a recent work outlining the standalone SCIANTIX (version 2.0) performance and its separate-effect validation database, we present its performance when coupled with fuel performance codes to simulate light water reactor fuel rods. The experiments selected for the comparative analysis constitute an initial integral validation database. The comparison focuses on conventional engineering quantities of interest, such as integral fission gas release, demonstrating the satisfactory performance of the code. Additionally, it highlights the potential advantages of multi-scale modelling over conventional semi-empirical approaches. • The integral validation database of SCIANTIX is detailed. • The code is coupled with the thermo-mechanical fuel performance codes TRANSURANUS, FRAPCON and OFFBEAT. • The validation database includes Light Water Reactor fuel rods operating in nominal operative and transient conditions. • The comparison is performed against engineering quantities such as the integral fission gas release. • Potential advantages of the multi-scale modelling and future developments are outlined. [ABSTRACT FROM AUTHOR]

• A bubble entrainment rate experiment based on WMS is completed. • The effect of different parameters on the bubble entrainment rate is analyzed. • A semi-empirical correlation for bubble entrainment rate is made. The fission gas in the thorium molten salt reactor's fission gas separation system is separated by creating a gas column in the pipe's center using a swirling flow field. During this process, if the swirling intensity fails to maintain the stability of interface, the gas column will become unstable and rupture, resulting in the formation of numerous entrained bubbles that could potentially impact separation efficiency. Thus, accurately estimating the entrainment rate of bubbles in the swirling flow field is crucial for predicting separation efficiency. This paper presents an experimental study on the bubble entrainment rate in a decaying swirling flow field using a wire-mesh sensor. The study analyzes the impact of various factors on the bubble entrainment rate and provides a semi-empirical correlation for the bubble entrainment rate in a decaying swirling flow field, with a prediction error of approximately ±30 %. [ABSTRACT FROM AUTHOR]

The current paper presents a review of key failure modes in AGR fuel, a description of failure mechanisms and the post failure effects that follow. The main failure modes of AGR fuel are found in the following proportions: ratchetting (axial clad straining and fuel-stacking with resulting fuel-stack gap formation and clad collapse) 3%; over pressure 4%; pellet clad mechanical interaction 7%; pin-brace interaction/rotating pins (fuel pin and support structure interaction, also known as fretting) 37%; clad restructuring/microstructural effects (including clad grain growth reducing ability of clad to withstand pellet clad interaction) 41%; other (including chipped pellets, endcap welds, and non-fuel component damage) 8%. A description of each failure mode is provided detailing the characteristics of the failure mode and a summary of the factors contributing to the failure mode. An overview of post-failure effects in AGR fuel is provided. Observations of failed fuel in hot-cells reveal a range of post-failure effects, in addition to the characteristics which contributed to the failure. Understanding post failure effects is important in order to isolate features which may have occurred prior to failure, and to inform on the behaviour of fuel between failure and discharge from reactor. [ABSTRACT FROM AUTHOR]

Metallic fuel undergoes dramatic microstructural changes early in life due to fission gas swelling until ∼2–3 at% burnup which affects the conductivity of the material, however the evolution of metallic fuel thermal conductivity during this early phase burnup has never been successfully measured in situ. The Irradiated Material Properties Accelerated Characterization Test (IMPACT) experiment will be the first in a series of experiments to irradiate advanced nuclear metallic fuel specimens with novel embedded thermal conductivity probes in ATR. In the current work the IMPACT experiment final design and supporting analysis is reported in detail. Results are evaluated for various reactor operational conditions to meet the functional requirements of the experiment. The first iteration of this IMPACT experiment will provide data regarding thermal properties evolution in uranium-zirconium (U10Zr) fuel, but this experiment vehicle is envisioned for future advanced fuels and structural materials irradiations in ATR. [ABSTRACT FROM AUTHOR]

The significant microstructural changes that U-Mo fuel undergoes during operation degrades its mechanical properties and structural integrity. Microstructural evolution entails the formation, evolution, and redistribution of porosity in conjunction with grain refinement. In the present paper, we employ numerical approaches to assess the impact of the various microstructural features—grains, nanoscale intragranular fission gas bubbles, and mesoscale intergranular voids—on the degradation of elastic constants. Phase-field microstructure models are combined with the asymptotic expansion homogenization technique in order to derive the effective elastic constants as a function of porosity and fission density. The results are verified and compared against theoretical bounds. Using this approach, elastic degradation in operating nuclear fuels can be quantified when the distributions of microstructural features are known. [ABSTRACT FROM AUTHOR]

Two tests, AN3 and AN4 from the RISOE-III Fission Gas Release (FGR) project are analysed using the updated version of the FALCON MOD01 code coupled with the GRSW-A model. Based on comparison of the stand-alone code calculation with the measurement for centerline fuel temperature, the previous calibration and verification of the thermal analysis of the code is confirmed. Additional analysis with the FALCON-to-FRELAX (F2F) coupled code system is carried out and presented for simulation of the effects of delayed axial redistribution of the released fission gases in the rodlets, as an attempt to interpret the data on significant jumps in measured upper-plenum pressure at a power dip. The other feature of the data, addressed by the additional sensitivity study with F2F, is peaking that can be seen in the measured fuel temperature after each power rise. A conclusion can be proposed that the peculiarities in the online-measurement could, probably, result from the features of fabrication of the fuel segments for pre-irradiation in the NPP, and instrumentation of the rodlets to be tested in the research reactor. Therefore, these features can have only limited effect on evaluation of reliability and safety of the standard fuel during normal operation in power reactors. [ABSTRACT FROM AUTHOR]

Cesium (Cs) is a radioactive fission product whose release is of concern for Tristructural-Isotropic fuel particles. In this work, Cs diffusion through high energy grain boundaries (HEGBs) of cubic- SiC is studied using an ab-initio based kinetic Monte Carlo (kMC) model. The HEGB environment was modeled as an amorphous SiC, and Cs defect energies were calculated using the density functional theory (DFT). From defect energies, it was suggested that the fastest diffusion mechanism is the diffusion of Cs interstitial in an amorphous SiC. The diffusion of Cs interstitial was simulated using a kMC model, based on the site and transition state energies sampled from the DFT. The Cs HEGB diffusion exhibited an Arrhenius type diffusion in the range of 1200-1600 °C. The comparison between HEGB results and the other studies suggests not only that the GB diffusion dominates the bulk diffusion but also that the HEGB is one of the fastest grain boundary paths for the Cs diffusion. The diffusion coefficients in HEGB are clearly a few orders of magnitude lower than the reported diffusion coefficients from in- and out-of-pile samples, suggesting that other contributions are responsible, such as radiation enhanced diffusion. [ABSTRACT FROM AUTHOR]

T91, a pivotal candidate material for cladding in lead-based fast reactors (LFRs), demonstrates considerable potential for corrosion resistance in liquid lead-bismuth eutectic (LBE) corrosion experiments. However, the multi-physics characteristics of T91 cladding corrosion in its operational environment remain unclear, posing an obstacle to its practical application. This study focuses on the spatiotemporal corrosion of T91 cladding in multi-physics coupling scenarios by means of COMMA (Cladding/Oxidation corrosion Multi-physics Modeling and Analysis code), in which, modules of cladding oxidation, cladding oxide removal, heat transfer, mechanical deformation, and fission gas release are integrated and collaboratively solved. The corrosion patterns of T91 cladding with different oxygen concentrations and the underlying multi-physics behavior of the corrosion processes under operational conditions simulating are summarized. For the lower oxygen concentration of 1.8e-9 wt% in LFRs, the results indicate a 0.2 m axial zone with base material exposed to LBE reveals the presence of the oxidation protection failure. An exacerbation of cladding heat transfer degradation is observed after oxygen concentration reaching the oxygen-dominant region, due to the enhanced temperature positive feedback of oxidation corrosion. For the high oxygen concentration of 1.4e-6 wt%, an axial offset of 0.24 m for the location of the minimum gap size is observed. This study identifies a trend of the reduction thickness of cladding that first decreasing and then increasing with an elevated oxygen concentration, attributed to a transition in corrosion patterns. The temperature and mechanical analyses indicate that the optimal oxygen concentration for corrosion inhibition and fuel safety is located around the critical oxygen-dominant concentration. • A multi-physics analysis program named COMMA is developed for investigating T91 corrosion behavior. • Oxidation corrosion of T91 cladding induces heat transfer degradation, leading to a shiftin practical oxygen control range. • For an oxygen concentration of 1.4e-6 wt%, non-uniformity corrosion causes a 0.24 m axial shift in minimum gap location. • Multi-physics analysis suggests that optimal oxygen concentration for T91 cladding is near oxygen-dominant concentration. [ABSTRACT FROM AUTHOR]

In this work, we quantify the impact of grain boundary (GB) and surface diffusion on fission gas bubble evolution and fission gas release in UO 2 nuclear fuel using simulations with a hybrid phase field/cluster dynamics model. We begin with a comprehensive literature review of uranium vacancy and xenon atom diffusivity in UO 2 through the bulk, along GBs, and along surfaces. In our model we represent fast GB and surface diffusion using a heterogeneous diffusivity that is a function of the order parameters that represent bubbles and grains. We find that the GB diffusivity directly impacts the rate of gas release via GB transport, and that the GB diffusivity is likely below 104 times the lower value from Olander and van Uffelen (2001). We also find that the surface diffusivity impacts bubble coalescence and mobility, and that the bubble surface diffusivity is likely below 10 − 4 times the value from Zhou and Olander (1984). [ABSTRACT FROM AUTHOR]

The post-irradiation anneal tests for dispersion fuel plates are commonly employed to determine the blistering threshold temperature in order to evaluate their resistance to accidents. In this study, a three-dimensional simulation method for UMo/Zr dispersion fuel plates is developed to predict the normal stresses in the fuel skeleton and some related variables during in-pile irradiation and out-of-pile anneal tests. The effects of the intrusion of extra fission gas atoms into gas bubbles are investigated, and the contributions of the creep performances of fuel particles and the matrix are examined. The predictions reveal the following insights: (1) the thermo-mechanical behaviors during anneal tests are strongly dependent on the intrusion of extra fission gas atoms into gas bubbles and the additional fission-gas induced swelling during post-irradiation anneal tests; the maximum tensile stress in the fuel skeleton increases significantly, from 57.4 MPa to 346.0 MPa, as the fraction of generated fission gas atoms in gas bubbles shifts from approximately 23.4 % to 100 % in the annealing process; (2) the thermo-mechanical responses during anneal tests are greatly influenced by the creep performances of fuel particles and the matrix; the significantly weakened creep ability of fuel particles is found to be the critical factors responsible for particle cracking in the annealing process; the fuel particle fracture will hardly occur when the irradiation creep rate model is maintained during the considered annealing scheme, because the skeleton tensile stress does not appear. It is important to note that the particle cracking and the subsequent blister behaviors under in-pile neutron irradiation conditions are possible to occur if a rapid temperature increase takes place under accident conditions. Consequently, the out-of-pile annealing tests are of significance. [ABSTRACT FROM AUTHOR]

• The anchoring ability of seventeen kinds of metal atoms on N 4 -G sheet was assessed. • The Sc-N 4 -G and Ti-N 4 -G nanographenes have been screened out as efficient adsorption substrate for Xe. • Hirshfeld charge of adsorbed Xe shows a strongest linear correlation with adsorption energy. It is highly important to develop the technology for the adsorption and separation of fission gases in the fields of nuclear environment monitoring and spent fuel disposal. The adsorption behaviors of fission gas Xe on a series of single-metal-atom doped N 4 -graphene (N 4 -G) were investigated with density functional theory (DFT) calculations in this work. The metal (M) atoms can be robustly anchored by the N-doped graphene (N 4 -G) in single-atom form rather than clustering together. The M-N bonds with the high-degree covalent nature account for the high binding energy of N 4 -G for single-metal-atom.The Sc-N 4 -G and Ti-N 4 -G with fairly low adsorption energy for Xe were screened out as efficient adsorbents from seventeen kinds of M-N 4 -G species studied. The high-degree charge redistribution in the Xe adsorption process is responsible for their high affinity for Xe. The Hirshfeld charge of adsorbed Xe on M-N 4 -G shows a high linear correlation with adsorption energy. Our study would provide useful insights into the potential applications of single-atom doped N 4 -graphene in the management of fission gases. A total of 17 single-atom doped on N 4 -G nanographene were screened for excellent Xe adsorption. [Display omitted] [ABSTRACT FROM AUTHOR]

Interface debonding may occur in the multilayered TRISO fuel particle and fully ceramic microencapsulated (FCM) dispersion nuclear fuels during their operation, exerting a great influence on their thermomechanical behaviors. In this study, a multiphysics coupling analysis on the performance of a TRISO particle and FCM fuel cell with the interface debonding effect considered is carried out. Interface debonding is dealt with the cohesive zone model (CZM) method. The implemented fuel performance models include temperature and burnup-dependent thermomechanical properties, gap heat conduction, thermal expansion, creep, irradiation swelling, fission gas production and release. The multiphysics coupling models and method are verified against International Atomic Energy Agency (IAEA) benchmarks. The effect of interface debonding between inner pyrolytic carbon (IPyC) layer and silicon carbide (SiC) layer on the performance of individual TRISO particle in traditional High Temperature Gas Reactors (HTGRs) is analyzed under zero traction boundary condition on the outer surface. The approach is then applied to simulate the effect of interface debonding between the SiC layer and outer pyrolytic carbon (OPyC) layer of the TRISO particle on the performance of a representative cell of the FCM fuel. It is shown that interface debonding can significantly affect the temperature distribution in the fuel and the stress magnitude in the SiC layer and matrix. The tensile stress in SiC layer in FCM fuel cell is unrealistically large when the interface bond strength is infinite, and the SiC/OPyC interface debonding reduces the tensile stress in the SiC layer to maintain its own structure strength, which provides the basis to reduce the SiC/OPyC interface bond strength in the manufacturing process of the FCM fuel. [ABSTRACT FROM AUTHOR]

Blistering is a unique failure form of plate fuel elements. During the long-term operation of the plate fuel element with dispersed fuel, the phenomenon of the local blister on the cladding surface caused by the accumulation of fission gas may occur. This phenomenon will not only affect the flow and heat transfer of coolant in the blister channel but also lead to a local high void fraction on the heating wall, and make the boiling crisis occur early. The forms and distributions of blisters on the cladding surface may be various under different conditions. At present, there are few reports on the distribution of void fraction in the blister channel under subcooled boiling conditions. However, CHF, which is one of the core thermal design criteria, is closely related to the distribution of void fraction. So, in this paper, the void fraction distribution in narrow rectangular channels with different blister heights and forms was studied by using the numerical simulation method with the Euler-Euler model and the modified RPI model. The calculation results showed that the void fraction at the back end of the blister area will have a local maximum, and the backflow of fluid at the back end of the blister is the main reason for the high void fraction. In addition, as the blister height increased, the local maximum value would also increase. When the blister height exceeded 1.5 mm, the local maximum void fraction would exceed 0.82, which would lead to a localized boiling crisis on the cladding surface. Under the same blister area and blister height conditions, multiple blisters were more likely to have multiple peak void fraction points than a single pillow blister. At the same location, the peak void fraction of blisters on both sides of the heating wall was higher than that on one side, and a boiling crisis might occur on both sides of the heating wall. [ABSTRACT FROM AUTHOR]

Irradiation with 84 MeV Xe ions at peak fluence above 7 × 1017 ions/cm2 has been carried out on two UO 2 samples with different initial grain sizes at 300 °C and 500 °C. Characterization of both grain structure and bubble morphology has been performed using transmission electron microscopy, image analysis and precession electron diffraction. The bubble size in the large-grained sample was smaller than that in the standard-grain sample, confirming fission gas precipitation resistance in the former. The ion irradiation in the high electronic energy deposition range has caused formation of a large density of dislocations which resulted in grain polygonization for the sample with standard grain size, which was observed in multiple lamellas extracted next to each other at peak dose. The sample with large grains showed inconsistent behavior. In one case, dislocation pile-up occurred, but no sub-grain formation was observed. An adjacent location showed sub-grain formation, even if to less extent compared to the standard grain-size sample. These findings suggest a limited polygonization resistance in large-grained UO 2 compared to standard-grain UO 2. Recrystallization of both sample surfaces was present, leaving unclear if this is a surface effect or a consequence of the highest energy deposition rate in this area. [ABSTRACT FROM AUTHOR]