Fission Gas Behavior Research Articles

Two-phase modelling for fission gas sweeping in restructuring nuclear oxide fuel

In this work, we propose a modelling approach for the intra-granular fission gas behaviour in UO2 under restructuring process. Leveraging the definition of restructured volume fraction, we consider the fuel matrix transition from the non-restructured to the restructured phase, together with the evolution of the corresponding fission gas concentrations retained in the fuel matrix. Firstly, we derive a sweeping term that exchanges fission gas atoms from the non-restructured to the restructured fuel region. The sweeping term is then included in the conventional intra-granular fission gas diffusion problem. Secondly, the spectral diffusion algorithm is employed to solve two spatially-dimensionless problems, properly representing the non-restructured region with micrometric grains and the restructured region with sub-micrometric grains. The model developed is implemented in SCIANTIX, a 0D meso-scale code for physics-based modelling of fission gas behaviour in nuclear oxide fuel and compared with experimental data and semi-empirical models.

CRUD deposition induced multi-physics effects analysis in PWR

In PWR, the deposition of corrosion products from the steam generator leads to the formation of Chalk River Unidentified Deposits (CRUD) on the surface of fuel rods. CRUD deposition can cause issues such as CRUD induced localized corrosion, axial power anomaly and deterioration of cladding heat transfer. A new multi-physics program named CIFF (CRUD Induced eFFects analysis) is developed to analyze the effects caused by CRUD deposits. CIFF is a 1.5D model and can be used to simulate the variations in surface temperature, cladding corrosion and fuel deformation caused by CRUD deposition. A three-node CRUD deposition model is developed based on mass transfer theory and chemical equilibrium. The calculations of heat transfer, mechanical deformation, oxide layer growth, CRUD deposition and fission gas behavior are fully coupled based on finite element analysis and numerical solution technology. Axisymmetric element is used to solve the deformation of fuel rod under thermal expansion, gas pressure, creep and radiation swelling. The performance of CIFF is validated through IFA-513, ANO-2, Oconee and PCCL cases. The DBN-1 case is used to explore the influence of CRUD. The results indicate that CRUD deposition and CRUD induced local corrosion deteriorate the heat transfer of cladding, resulting in a three-stage increase of cladding temperature and accelerated growth of the oxide layer. CRUD deposition also leads to a relative increase in the gap pressure exceeding the coolant pressure by 8.4 %, a relative increase in the fuel rod strain by 10 % and a relative decrease in the gap size by 11 %, which indirectly leads a premature pellet-cladding mechanical interaction by 90 days.

Phase-field simulation on fission gas release behavior of large grain UO<sub>2</sub> fuel

In order to predict the release behavior of fission gas in large grain UO<sub>2</sub> fuel and provide support for the development of accident tolerant fuel, a phase-field model is used to simulate the release behavior of fission gas in the microstructure of UO<sub>2</sub> polycrystalline in this work. This model adopts a set of coupled Cahn-Hilliard equations and Allen-Cahn equations, using conserved field variables to represent the distribution of fission gas and vacancies, and distinguishing bubble phase from matrix phase by using order parameters. This model focuses on investigating the effects of different grain sizes, temperature conditions, and diffusion coefficients on the release behavior of fission gas, demonstrating the nucleation, growth, and fusion behavior of bubbles. Simulation results are obtained for fuel porosity, bubble coverage on grain boundaries, and average bubble radius at a certain degree of burnup. The results show that temperature and diffusion coefficient have a significant influence on porosity and bubble coverage on grain boundaries. When the diffusion coefficient is high, grain size also has a significant influence on fission gas release behavior. And when the diffusion coefficient is low, the influence of grain size is not significant. In addition, the distribution of fission gas bubbles under high burnup obtained through this model is also in good agreement with experimental result. The model can predict the behavior of fission gas release in large grain UO<sub>2</sub> fuel.

Study of xenon evolution in UO[formula omitted] using multi-grain cluster dynamics modeling

The behavior of fission gas is a key issue for the performance of UO2 fuel, which is the most widely used fuel type in GEN-III reactors. Fission gas will lead to lower thermal conductivity and affect the microstructure of fuel, gradually cause swelling and other phenomena. It seriously affects the performance and safety of nuclear reactor. The numerical simulation method based on physical model is an important means to study fission gas. In this paper, in order to study the key parameters affecting the behavior and evolution of fission gas xenon in UO2 fuel, a multi-grain parallel cluster dynamics model is designed and implemented. The model can describe the multi-grain structure characteristics of UO2, and establish the correlation for the physical process of intra- and inter-granular bubbles. Furthermore, the model is extended to a larger temporal and spacial scale through parallelization, so as to simulate the fission gas behavior closer to the actual situation, and expand the applicable conditions and range of the model. Based on the modeling work, the long-term evolution processes of xenon under different initial grain radius and temperature conditions are analyzed, the results show that the initial grain radius has a much bigger effect on the inter-granular gas concentration than the intra-granular gas, while temperature has a large effect on both intra- and inter-granular gas concentrations. The model and analysis results provide theoretical support for understanding fission gas behavior and improving UO2 fuel performance.

The SCIANTIX code for fission gas behaviour: Status, upgrades, separate-effect validation, and future developments

SCIANTIX is a 0D, open-source code designed to model inert gas behaviour within nuclear fuel at the scale of the grain. The code predominantly employs mechanistic approaches based on kinetic rate-theory models to calculate engineering quantities, such as fission gas release and gaseous fuel swelling. Since its release, SCIANTIX has undergone significant improvements, including the incorporation of new modelling and numerical capabilities. The code architecture has been revamped, embracing an object-orientated structure improving the overall efficiency and usability. This work provides a concise overview of the current state of the SCIANTIX code, highlighting recent updates and advancements. Each SCIANTIX model is presented along with the corresponding separate-effect validation database, which is used to assess its accuracy and predictions.

Extended Development of a Fission Gas Release Behavior Model Inside Spherical Fuel Grains for LWR Reactors

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.

A reduced order model for fission gas diffusion in columnar grains

In fast reactors, restructuring of the fuel micro-structure driven by high temperature and high temperature gradient can cause the formation of columnar grains. The non-spheroidal shape and the non-uniform temperature field in such columnar grains implies that standard models for fission gas diffusion can not be applied. To tackle this issue, we present a reduced order model for the fission gas diffusion process which is applicable in different geometries and with non-uniform temperature fields, maintaining a computational requirement in line with its application in fuel performance codes. This innovative application of reduced order models as meso-scale tools within fuel performance codes represents a first-of-a-kind achievement that can be extended beyond fission gas behaviour.

A modelling methodology for the description of helium behaviour accounting for the grain-size distribution

Physics-based meso-scale models of fission gas behaviour for fuel performance codes currently consider only the average grain size as physical parameter to describe the fuel microstructure. Nevertheless, information on the grain-size distribution is available for several metallographically characterized fuel samples. To extend the current modelling approach, we present new experimental data and develop a methodology to treat the fuel grain-size distribution based on multi-grain analysis. This is applied to describe helium behaviour from infused and annealed polycrystalline UO2 samples. The methodology consists of three steps: (1) Acquisition of the empirical grain distribution from sections of the polycrystals, (2) Calculation of the gas distribution factor pertaining to each grain size class after helium infusion, and (3) Simulation of the experimental annealing histories with input parameters weighted by the gas distribution factors. To perform the multi-grain analysis, we used the SCIANTIX code, which allows calculating the helium kinetics in a single fuel grain. The application of this methodology is promising because it can represent the dynamics of helium release more consistently with the physics of the fuel microstructure compared to the state-of-art approaches, and it can be suitable for application in fuel performance codes for different enhanced accident tolerant fuel concepts as outlined in our conclusions.

Modeling of fission gas diffusion and release for doped

Uranium dioxide (▪) is the primary nuclear fuel in light water reactors, and its excess neutronic reactivity can be controlled by adding burnable absorbers, such as ▪. This burnable absorber has a large neutron absorption cross-section, lowering the high reactivity of the reactor's initial fuel load. However, there needs to be more understanding of how added ▪ influences the properties of ▪ under irradiation. To understand the behavior of defects and fission gas in the ▪/▪ system under irradiation, we use cluster dynamics modeling supported by density functional theory calculations. First, we calculate the formation energies of Gd point and cluster defects, and evaluate the temperature-dependent defect concentrations using the defect formation energies and entropies. We show that Gd is soluble in ▪, introducing a negative charge in the system. Using this information, we adapted the cluster dynamics code Centipede to model the influence of Gd on U self-diffusion and Xe diffusion in ▪ with 10 wt% ▪. Also, we analyzed the Xe diffusion as a function of ▪ concentration, showing that the Xe diffusivity is decreased, which means that the athermal diffusivity due to electronic stopping persists at higher temperatures. The decrease in Xe diffusion means that more Xe stays in the matrix, decreasing the Xe release, and lowering its influence of fission gas release on the thermomechanical properties of ▪.

Integrated simulation of U-10Mo monolithic fuel swelling behavior

A separate computational branch has been implemented within the DART (Dispersion Analysis Research Tool) computational code to simulate the swelling behavior of U-10Mo monolithic fuel under the operating conditions of high-power research and test reactors (RTRs). The monolithic branch of the DART code implements a mechanistic rate-theory-based fission-gas-behavior model for the calculation of fission gas swelling, as well as a suite of thermal, physical, and mechanical models to consider various processes occurring in RTR fuels during irradiation. To accurately simulate and eventually predict U-10Mo monolithic fuel irradiation behavior, the code uses materials properties calculated with lower length-scale computational methods, such as gas atom diffusivity and U-Mo surface energy from atomic simulations and grain-morphology-specific recrystallization kinetics (recrystallized fuel volume fractions versus fission density) predicted using the phase-field method. The remainder of the fission gas behavior parameters used in the model were calibrated with measured intergranular bubble size distributions. With this integrated simulation approach, the swelling behavior of U-10Mo monolithic fuel was simulated for various initial grain sizes at different operating conditions and compared with measured data. Furthermore, because limited experimental data exist for parameter calibration, detailed sensitivity studies for the important parameters used in the fission gas behavior model were performed to examine their impact on both intergranular gas bubble morphology at low fission density, and on total porosity at high fission density.

Diffusion of Xe and Kr implanted at low concentrations in UO2 as a function of temperature – An experimental study

Fission gases production and release have a large impact on uranium dioxide fuel performance. To predict fuel properties in pile, a better understanding of the fission gas behaviour in uranium dioxide is needed. UO2 samples were implanted with Xe or Kr at low doses to avoid trapping by irradiation defects and/or gas clusters. The release rates of Xe and Kr were studied at temperatures ranging up to 1400°C. Results show an excellent agreement with the substantial literature on xenon diffusion in irradiated UO2.

Accelerating cluster dynamics simulation of fission gas behavior in nuclear fuel on deep computing unit–based heterogeneous architecture supercomputer

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.

Post-irradiation examination of low burnup U3Si5 and UN-U3Si5 composite fuels

This work presents post-irradiation examination data on UN-U3Si5 and U3Si5 fuels at low burnup (i.e., <10–15 GWd/tHM) with Kanthal AF® cladding. The results suggest good irradiation performance for both the silicide and nitride-silicide composite pellets. Optical microscopy revealed that the pellet-cladding gap is still open, and limited axial cracking was observed only in UN-U3Si5 pellets. Microcracking was isolated to the U3Si5 phase in all cases and was observed in pre-irradiation and depleted pellets, indicating that it was not irradiation induced. The fission gas release was minimal for the calculated fission density achieved (2.6 – 3.15 × 1020 fiss/cm3). No fission gas bubbles were observed in the optical metallography. These results suggest acceptable swelling and fission gas behavior for both the single phase and composite compositions.

Performance evaluation of thermal conductivity models for fission gases with application to [formula omitted]-zirconium dispersion fuel

In this paper, the modeling of fission gases, particularly xenon and krypton, and their effects on the thermal performance with application to UO2-zirconium dispersion fuel is investigated.To investigate the consequence of fission gas behavior in nuclear dispersion fuel, in the first step, our obtained single atom diffusion coefficient, Ds, of xenon in UO2 by implementing molecular dynamics is used and open source code SCIANTIX is improved and employed to obtain the density of bubbles through the fuel. A microstructure model of dispersed fuel has been used to treat fuel meat thermal conductivity by taking into account fission gas bubbles density. Furthermore, the results of other models are given to evaluate fission bubbles' evolution.To explore thermal conductivity models of dispersion fuel, a code called Kowsar-1 is developed by including fission gases generation.

First principles investigation of trapping and diffusion of fission gas in AnO2

The trapping of fission gases in AnO2 has been investigated using density functional theory. The behavior of fission gases varies with the size in these oxides, therefore He and Xe have been considered in this work. In ThO2 and UO2, the octahedral interstitial site has been found to be more stable than oxygen vacancy, while actinide vacancy is found most stable site for He in all three oxides. For Xe, an oxygen vacancy is a more stable site than an octahedral interstitial site and actinide vacancy is the most stable site in these oxides. Our results indicate that He is much more mobile than Xe in these oxides.[copyright information to be updated in production process]

The effect of interfacial phenomena on gas solubility measurements in molten salts

The behavior of fission gases in molten fuel salt reactors governs activity transport from the reactor and can also affect the performance of the reactor itself. The gas solubility can be described thermodynamically by Henry’s law. However, the coupling of the condensed and gas phases depends on the interfacial area, which is difficult to measure or even to estimate. Surfaces of materials in the reactor will include disperse phases in the salt and porosity within the structural materials, covering a range of compositions and sizes. These attributes can affect measurements of fundamental properties such as gas solubility. Methods to obtain gas solubility, surface tension, interfacial energies, and bubble gas transport are reviewed. Recent data from manometric experiments are interpreted based on xenon sorption onto salt-wetted quartz.

Assessment of INSPYRE-extended fuel performance codes against the SUPERFACT-1 fast reactor irradiation experiment

Design and safety assessment of fuel pins for application in innovative Generation IV fast reactors calls for a dedicated nuclear fuel modelling and for the extension of the fuel performance code capabilities to the envisaged materials and irradiation conditions. In the INSPYRE Project, comprehensive and physics-based models for the thermal-mechanical properties of U–Pu mixed-oxide (MOX) fuels and for fission gas behaviour were developed and implemented in the European fuel performance codes GERMINAL, MACROS and TRANSURANUS. As a follow-up to the assessment of the reference code versions (“pre-INSPYRE”, NET 53 (2021) 3367–3378), this work presents the integral validation and benchmark of the code versions extended in INSPYRE (“post-INSPYRE”) against two pins from the SUPERFACT-1 fast reactor irradiation experiment. The post-INSPYRE simulation results are compared to the available integral and local data from post-irradiation examinations, and benchmarked on the evolution during irradiation of quantities of engineering interest (e.g., fuel central temperature, fission gas release). The comparison with the pre-INSPYRE results is reported to evaluate the impact of the novel models on the predicted pin performance. The outcome represents a step forward towards the description of fuel behaviour in fast reactor irradiation conditions, and allows the identification of the main remaining gaps.

Influence of point defect accumulation on in-pile thermal conductivity degradation: Fuel rod defect distribution and deviation between in-pile and post irradiation thermal conductivity

As nuclear fuel burn-up increases, its thermal conductivity degrades due to the accumulation of defects that lead to increased phonon scattering rates. This results in a rise in the centerline temperature of the fuel rod, whereby heat generation must be decreased to avoid undesired behavior such as fuel melting and extensive fission gas release. Fuel performance codes are utilized to optimize the fuel's operational conditions; while they are based on established physical principles, their empirical nature limits their predictive capabilities. Recently, an effort has been made to develop predictive fuel performance codes for commonly used nuclear fuels, as well as accelerated qualification of advanced nuclear fuels. In this report, we elaborate on the importance of careful analysis of point defects’ impact on thermal conductivity in fuel performance analysis. A model is presented where point defect concentration is estimated based on Rate Theory modeling and used as input to the Klemens-Callaway model to calculate their contribution to the degradation of thermal conductivity in UO2 under prototypical irradiation conditions. This analysis suggests that point defect concentration is significant at the rim of the fuel pellet and neglecting this leads to underestimation of the centerline temperature, which may have consequences on fission gas behavior.

Towards grain-scale modelling of the release of radioactive fission gas from oxide fuel. Part II: Coupling SCIANTIX with TRANSURANUS

The behaviour of the fission gas plays an important role in the fuel rod performance. In a previous work, we presented a physics-based model describing intra- and inter-granular behaviour of radioactive fission gas. The model was implemented in SCIANTIX, a mesoscale module for fission gas behaviour, and assessed against the CONTACT 1 irradiation experiment. In this work, we present the multi-scale coupling between the TRANSURANUS fuel performance code and SCIANTIX, used as mechanistic module for stable and radioactive fission gas behaviour. We exploit the coupled code version to reproduce two integral irradiation experiments involving standard fuel rod segments in steady-state operation (CONTACT 1) and during successive power transients (HATAC C2). The simulation results demonstrate the predictive capabilities of the code coupling and contribute to the integral validation of the models implemented in SCIANTIX.

Numerical study on the release and migration behavior of fission gas in a molten LBE pool

The lead-cooled fast reactor (LFR) is one of the most promising fast neutron reactors using molten lead or the lead–bismuth eutectic (LBE) alloy as a coolant. Under postulated severe accidents, the fuel rod of LFR may be damaged, which would cause the release of fission gas, and the migration of fission gas bubbles in the reactor molten pool will affect the release and absorption of radioactive substances in the reactor. In this paper, a three-dimensional numerical study on the release and migration behavior of fission gas in the molten LBE pool of LFR is carried out based on the volume of fluid method. The bubbles are continuously released by gas injection, and the research mainly focuses on the detachment time, the rising velocity, and the size of the bubble when it detaches at the orifice. The coalescence of bubbles is observed, and the acceleration effect of the bubble wake is confirmed. The distribution of the bubble terminal rising velocity with diameter has no simple or linear relationship. The effects of the gas injection velocity, the release depth, and the gas injection angle are studied. A lower gas injection velocity will delay the detachment and reduce the size of the bubble. The increase of release depth tends to release smaller bubbles. The bubbles released from a vertical surface will attach to the wall. The simulations and theoretical analysis are comparable and have similar tendencies. The distribution of the bubble terminal rising velocity with equivalent diameter may predict the migration behavior of bubbles in molten LBE.