Fission Gas Release Model Research Articles

An integrated statistical-thermodynamic model for fission gas release and swelling in nuclear fuels

We propose a new model for burst fission gas release induced by microcracking in ceramic nuclear fuels such as uranium dioxide. The model stipulates that the densities of defects in the fuel material, such as microcracks and fission gas bubbles on grain boundaries, evolve in accordance with the second law of thermodynamics. Central to the model is the notion of an effective temperature, conjugate to the configurational entropy of the fuel material, and directly linked to the burnup. The model predicts that microcracking, driven by the internal stress state of the fuel material, reduces the bubble storage capacity of grain boundaries, and accounts for burst fission gas release during rapid temperature transients that simulate power transients, reactor startup, and loss-of-coolant accident conditions.

Effects of transient fission gas release on rod balloon burst behavior during a loss-of-coolant accident

The US nuclear energy industry is investigating strategies to increase the reactor operating cycle to 24 months, which would result in rod average burnups exceeding the current limit of 62 GWd/tU. To support this goal,multiphysics simulations tools and methodologies are being developed to predict the effects of this change duringtransient events such as loss-of-coolant accidents (LOCAs). In this work, two cladding burst correlations were used to predict cladding failure. These were coupled with three transient fission gas release (tFGR) models and were implemented in the BISON fuel performance code to quantify any changes in cladding burst behavior during a large-break LOCA. First, a simple linear model was used to perform a sensitivity analysis of a single fuel rod. Second, a tFGR model available in BISON was applied to 281 fuel rods throughout the core. The third model examined was an empirical correlation based on data from the literature.

Upgradation of fission gas release model in the nuclear fuel analysis code FuSPAC

The evaluation of fuel performances in the whole life cycle is important to the fuel integrity. In our previous work, a fuel performance analysis code FuSPAC was developed and validated against the in-pile test data and FRAPCON code. In the current work, this code was upgraded by replacing the modified Forsberg-Massih for fission gas release (FGR) with a mechanism model. The microscopic behaviors related to the FGR, including the gas atom diffusion, trapping, re-solution and bubble nucleation inside the granular particles, the vacancy diffusion at the inter-granular boundaries, the grain growth and the grain recrystallization were modeled. The upgraded FuSPAC code was validated against the in-pile test data for FGR and internal void volume. The sensitivity analysis on the sub-models shown that the intra-granular fission gas diffusion, the re-solution, and the inter-granular vacancy diffusion had significant effects on the FGR. The effects from rod power can be well reflected by the upgraded code compared to the original code.

Empirical and mechanistic transient fission gas release model for high-burnup LOCA conditions

In response to the nuclear industry desire to extend burnup beyond current licensing practices, the US Nuclear Regulatory Commission (NRC) released a research information letter (RIL) that provides a basis for analyzing fuel fragmentation, relocation, and dispersal (FFRD) in light-water reactors. Of the five elements discussed, the most ambiguous is the significance of transient fission gas (FG) release (FGR) (tFGR) and its effects on fuel performance under loss-of-coolant accident conditions. In fresh fuel, FG migration and eventual release is primarily governed by diffusion-based mechanisms at higher temperatures (>1,000 °C). However, the mechanisms governing FGR changes as burnup increases. More recent research indicates that FGR increases as burnup increases, specifically under temperature transient conditions, and this release occurs at lower temperatures with a new release mechanism. This behavior has been attributed to microcracking in the fuel and is likely related to microstructure embrittlement with the presence of over pressurized FG bubbles. The NRC RIL outlines the complexity of the phenomenon and a need for a deeper understanding to adequately address FFRD for regulatory application. Therefore, this manuscript intends to summarize the publicly available tFGR data and discuss the observed dependencies (e.g., burnup, heating rate, sample geometry, terminal temperature). An empirical model has been developed and benchmarked against recently published experimental data. However, this empirical model is limited to conditions for which fitting data exist and, therefore, a high-level discussion is included to provide a roadmap for atomistically-informed multiscale modeling in conjunction with experimental data collection to develop a mechanistic tFGR model widely applicable to a broad range of nuclear fuel conditions at high burnup.

Nuclear Fuel Modelling During Power Ramp

Fuel rods operating for several years in a LWR can experience fuel-cladding gap closure as a result of the phenomena due to temperature and irradiation. Local power increase induces circumferential stresses in the cladding as a result of the different expansion in the cladding and the pellet. In presence of corrosive fission products (i.e. Iodine) and beyond specific stress threshold and level of burnup, cracks may grow-up from the internal to the external cladding surface, causing fuel rod failure. The phenomenon, known as pellet cladding interaction-stress corrosion cracking PCI/SCC, or PCI, has been identified as a problem since the 70's. The PWR Super-Ramp experiment (part of OECD/NEA “International Fuel Performance Experiments (IFPE) database”) twenty eight fuel rods behaviour has been simulated using TRANSURANUS code version “v1m1j11”. Two sets (“Reference” and “Improved”) of suitable input decks modelling the fuel rods, based on the available literature are used to run the simulations. Focus is given to the main phenomena which are involved or may influence the cladding failure. Systematic comparison of the code results with the experimental data are performed for the parameters relevant for the PCI phenomenon. Sensitivity calculations on fission gas release models implemented in TRANSURANUS code are also performed in order to address the impact on the results. The results show the ability of TRANSURANUS version “v1m1j11” in conservatively predicting the rods failure due to PCI in PWR fuel and Zircaloy-4 cladding. Increased availability of experimental data would help to perform a deeper analysis.

Modeling of fission product release during severe accidents with the fuel performance code ALCYONE

This paper presents simulations of four tests performed on medium to high burnup fuel during the VERCORS and VERDON experimental programs. The tests are representative of a Severe Accident (SA) sequence with a temperature increase up to fuel-clad melting and oxidizing/reducing conditions within the furnace. The simulations are performed with the fuel performance code ALCYONE where irradiated fuel thermochemistry and fission gas release are coupled. In this paper, the impact of the radial burnup and Fission Product (FP) profiles within the fuel pellet on the FP release from the sample during the SA sequence is studied. Simulations of the fuel sample behavior during nominal irradiation in commercial reactors are first performed to assess the initial state of the fuel. The simulations of the SA sequences include a burnup dependent fission gas release model. The simulated release curves of various volatile and semi-volatile FPs (Xe, I, Te, Cs, Mo and Ba) are compared successfully to on-line measurements. The impact of the burnup and FP radial profiles on both the thermochemical equilibria within the pellet and the FP release kinetics is discussed. It is shown that the FP release from the fuel pellets is not significantly increased by the consideration of the burnup and FP radial profiles. This conclusion is due to the limited radial extension of the peaked burnup radial profile in the fuel pellet and to the uniform temperature of the fuel samples.

Study on the effect of long-term high temperature irradiation on TRISO fuel

In the core of the WWR-K reactor, a long-term irradiation of tristructural isotopic (TRISO)-coated fuel particles (CFPs) with a UO2 kernel was carried out under high-temperature gas-cooled reactor (HTGR)-like operating conditions. The temperature of this TRISO fuel during irradiation varied in the range of 950–1100 °C. A fission per initial metal atom (FIMA) of uranium burnup of 9.9% was reached. The release of gaseous fission products was measured in-pile. The release-to-birth ratio (R/B) for the fission product isotopes was calculated. Aspects of fuel safety while achieving deep fuel burnup are important and relevant, including maintaining the integrity of the fuel coatings. The main mechanisms of fuel failure are kernel migration, silicon carbide corrosion by palladium, and gas pressure increase inside the CFP. The formation of gaseous fission products and carbon monoxide leads to an increase in the internal pressure in the CFP, which is a dominant failure mechanism of the coatings under this level of burnup. Irradiated fuel compacts were subjected to electric dissociation to isolate the CFPs from the fuel compacts. In addition, nondestructive methods, such as X-ray radiography and gamma spectrometry, were used. The predicted R/B ratio was evaluated using the fission gas release model developed in the high-temperature test reactor (HTTR) project. In the model, both the through-coatings of failed CFPs and as-fabricated uranium contamination were assumed to be sources of the fission gas. The obtained R/B ratio for gaseous fission products allows the finalization and validation of the model for the release of fission products from the CFPs and fuel compacts. The success of the integrity of TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated. A low fuel failure fraction and R/B ratios indicated good performance and reliability of the studied TRISO fuel.

Modeling high burnup fuel thermochemistry, fission product release and fuel melting during the VERDON 1 and RT6 tests

This paper presents simulations of the VERDON 1 and RT6 tests (temperature increase up to fuel-clad melting, oxidizing and/or reducing conditions within the furnace) performed with high burnup UO2 fuel (i.e., up to 72 GWd/tU) and considering a coupling between irradiated fuel thermochemistry and a fission gas release model. The thermochemical calculations rely on the Thermodynamics of Advanced Fuels - International Database (TAF-ID) for the description of the phases likely to form from the 15 fission product considered in the fuel (Ba, Ce, Cs, I, La, Mo, Nd, Np, Pd, Rh, Ru, Sr, Tc, Te, Zr) and on the OpenCalphad solver for the minimization of the Gibbs energy of the system. The gas release model describes a diffusion process for the gases within equivalent spherical grains. It accounts for the gases formed by reaction between the fission products as well as the chemically inert noble gases. The progressive fission product depletion of the fuel due to their release in gas form closes the coupling. The impact of the fission product release kinetics on the thermochemical equilibria within the fuel is then studied by coupling the thermochemical calculations with a gas diffusion model. The coupled simulations led to a very good agreement with the release kinetics of various fission products (I, Te, Cs, Mo, Ba). The released fractions of the low-/non-volatile fission products and of uranium, plutonium are also well reproduced. The observed differences in the fuel-clad melting temperatures in the two tests (2200 °C during RT6 and 2600 °C during VERDON 1) are however not reproduced by the simulations.

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

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

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

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

Three-dimensional reconstruction from experimental two-dimensional images: Application to irradiated metallic fuel

This work applies reconstruction methods based on a genetic algorithm to derive 3D material properties, namely porosity and percolation fraction, in irradiated U-Pu-Zr fuel with minor actinides. We provide two-dimensional experimental data regarding the radial distribution of fission gas bubbles in the fuel and apply the algorithm successfully developed in a companion paper to reconstruct the fuel pore structure in 3D which is unknown a priori. The algorithm returned a set of best structures that constituted the best candidate solutions representing the pore phase. From these, it was possible to extract statistics on the 3D percolation fraction of the reference medium and infer a mean value, the related uncertainty, and an upper and lower bound of the percolation fraction. The algorithm proved able to infer this 3D property from 2D information of the metallic fuel with confidence intervals, thus establishing a path to infer 3D properties directly from 2D experimental images. The knowledge of such a relationship can be used to extrapolate the percolation threshold with confidence interval, which is a crucial property in defining microstructure-based fission gas release models of metallic fuels.

TRISO particle fuel performance and failure analysis with BISON

Because of its widespread use in multiple advanced reactor concepts, the ability to accurately simulate tri-structural isotropic (TRISO) fuel performance is essential for ensuring the safe operation of these reactors. To that end, the BISON fuel performance code has undergone significant recent development to expand its TRISO particle fuel performance modeling capabilities. This includes the development of material models, such as elastic, creep, swelling, thermal expansion, thermal conductivity, and fission gas release models. The capability to perform statistical failure analysis on large sets of samples has also been developed, utilizing a Monte Carlo scheme to execute fast-running 1-D spherically symmetric models. Stress adjustments are made in those 1-D models to account for multi-dimensional failure phenomena. Stress correlation functions are extracted from multi-dimensional failure simulation results, such as from a particle with cracked inner pyrolytic carbon (IPyC) and an aspherical particle. This paper provides a detailed description of the models used by BISON for TRISO fuel, along with a set of problems that test these models by (favorably) comparing them both with another code and experimental data. These include simulations of the Advanced Gas Reactor (AGR)-2 and AGR-5/6/7 experiments, with predictions for fuel performance parameters, failure probability, and fission product transport.

IMPLEMENTATION AND VALIDATION OF A FISSION GAS RELEASE MODEL FOR CTFFUEL USING THE NEA/OECD IFPE DATABASE

The nuclear reactors themselves are complex systems whose responses are driven by interactions between different physics phenomena within the reactor core. Traditionally, the different physics phenomena have been analyzed separately and its interaction considered via boundary conditions or closure models. However, in parallel with the development of computational technology, multi-physics coupled simulations are being used to obtain accurate predictions thanks to the consideration of the feedback effects on the fly (on-line). In the nuclear systems the fuel temperature is an important feedback parameter used to obtain the nuclear cross sections at given conditions by the neutron kinetics codes. An accurate prediction of temperature profile within the fuel rod involve several physics such as neutron kinetics, mechanics, material behavior and properties, heat transfer, thermal-hydraulics, and even chemistry. The pellet to clad gap conductance is possibly the most important source of uncertainty in the solution of conductivity equation in the fuel rod and the fuel temperature prediction. The gap conductance depends on two effects: the pellet to gap distance and the conductivity of the gas species that fill the gap. In this research work, the authors are focused on improving of the prediction of the gap gas conductivity in CTFFuel by implementing a fission gas release model in the code. The objective of this contribution is the implementation of a transient fission gas release model in CTFFuel and its validation using the experimental data available in the OECD/NEA International Fuel Performance Experiments (IFPE) database. CTFFuel is an isolated fuel heat transfer capability within the framework of CTF code, the state-of-the-art version of the Coolant Boiling in Rod Arrays Code – Two-Fluid (COBRA-TF) sub-channel thermal-hydraulic code. The code is being jointly developed by North Carolina State University (NCSU) and Oak Ridge National Laboratory (ORNL) within the US Department of Energy (DOE) Consortium for Advanced Simulation of LWRs (CASL).

Application of Kriging and Variational Bayesian Monte Carlo method for improved prediction of doped UO2 fission gas release

One of the advanced nuclear fuel concepts for current commercial water-cooled reactors focuses on microstructural modification of UO2 fuel via dopants. Dopants can effectively promote grain growth and suppress fission gas release (FGR), a key parameter that dictates the overall nuclear fuel performance. This work improves the BISON FGR model for chromia/alumina-doped UO2 fuel through statistical calibration with in-reactor experimental data. The high computing cost and nonintrusive nature of BISON limit the application of conventional techniques under the Bayesian framework. Dimensionality reduction is performed using principal component analysis (PCA) to deal with the FGR time series data. Kriging is used as metamodel of BISON to reduce the computing cost. A novel optimization framework, Variational Bayesian Monte Carlo (VBMC) is demonstrated as a low-cost nonintrusive approach for Bayesian calibration. The performance of VBMC is compared to the conventional statistical Markov Chain Monte Carlo (MCMC) sampling showing similar accuracy but superior efficiency.

Multidimensional multiphysics modeling of TRISO particle fuel with SiC/ZrC coating using modified fission gas release model

TRISO particle is adopted in the design of various types of reactors and accident tolerant fuel (ATF) due to its excellent accident tolerant performance under high temperature and other severe conditions. This study presents a multidimensional thermomechanical analysis of TRISO particle fuel with SiC/ZrC coating under typical operation conditions in reactor. Key fuel performance phenomena were considered including thermal performance (heat generation, conduction, irradiation gap heat transfer, etc.), mechanics (thermal expansion, elastic strain, densification, fission product swelling strain, contact, and thermal, irradiation creep, etc.), also irradiation behaviors (burnup, fission gas production and release, gap/plenum pressure with plenum volume, etc.). All these models have been integrated and solved via finite-element method. As for results, several complicated models were implemented and proved to successfully predict the thermomechanical behaviors of TRISO coated fuel compared with the existing experimental or numerical results. An enhanced fission gas behavior model considering grain boundary diffusion was established, which could complete detailed calculation on fission gas release at both grain size and fuel particle pellet scales. Based on the preliminary thermomechanical analysis, in terms of structural strength and heat transfer efficiency, ZrC appears to be a candidate material of SiC. Moreover, failure mechanisms and effective thermal conductivity have been analyzed using the current modified analysis technique.

Development of a thermo-mechanical coupling code based on an annular fuel sub-channel code

To simulate the thermal hydraulic performance of the annular fuel more accurately, some mechanical models have been incorporated into the sub-channel code SACAF, including pellet displacement model, gap heat transfer model, fission gas release model and fuel-cladding mechanical interaction model. The results calculated by the developed code are in good agreement with that by FROBA-ANNULAR. The developed code can consider the change of the heat transfer coefficient of the inner and outer gaps, the release of the fission gas release, the inner pressure of the fuel, the stress and strain of the cladding, which cannot be calculated by the conventional sub-channel code. Based on the calculation results, a great change is found in the gap heat transfer coefficient and MDNBR during the operation, which implies that the coupling of the sub-channel code and the mechanical performance models is necessary.

Development of an equilibrium loading pattern and whole-core fuel performance assessment in the Advanced Boiling Water Reactor (ABWR) with UO2 and U3Si2 fuels

The Advanced Boiling Water Reactor (ABWR) is an established evolutionary water reactor that has successfully achieved design certification in a number of countries. However, core design information is scarce in the open literature and this makes studying the suitability of new fuel types difficult and complicates comparisons with other reactor systems. This study therefore aimed to address important data gaps in the open literature relating to the latest generation of BWR core designs. Furthermore, this paper describes an assessment which has been carried out comparing the performance of U3Si2 and standard UO2 fuel deployed in the ABWR.An ABWR core design model has been created using the reactor physics software package CASMO-SIMULATE, with the method used to develop the model and the key set of parameters employed to determine the viability of the core design detailed. A major part of the optimisation process involves the application of simulated annealing to generate loading pattern candidates. Whilst simulated annealing has been employed routinely in loading pattern optimisation, one of the novel aspects in this study relates to each candidate adhering to a predetermined batch map to develop a viable equilibrium loading pattern.NNL's whole-core fuel performance framework NEXUS, which utilises ENIGMA as the fuel performance engine, was used in conjunction with the developed core model to investigate how standard UO2 fuel behaves in a modern BWR design and therefore highlight the most limiting fuel performance characteristics. The NEXUS results highlighted that, for the core designed here, the most limiting fuel performance characteristic for UO2 fuel was rod internal pressure.In addition, a first preliminary whole-core fuel performance assessment was carried out for the proposed enhanced Accident Tolerant Fuel candidate U3Si2 using the same power histories from the UO2 core design to enable a comparison of fuel behaviour independent of power history discrepancies. The assessments for U3Si2 have focused largely on temperature as, due to lack of U3Si2 in-pile experimental measurements, fuel performance parameters that are heavily dependent on swelling and fission gas release models currently have large uncertainties regarding their validity. Therefore, we begin to quantify the potential temperature characteristics associated with U3Si2 relative to UO2, in particular the peak homologous fuel temperature (that is the maximum fuel temperature throughout irradiation as a fraction of its melting point) for U3Si2, which was found to be 0.51 compared with 0.61 for UO2 fuel operating under the same conditions. As a further comparison between UO2 and U3Si2, simple single pin power-to-melt calculations were performed using ENIGMA. It was found that power-to-melt for UO2 is around 88 kW/m and the power-to-melt for U3Si2 is around 230 kW/m.

Phase-field modeling of fission gas bubble growth on grain boundaries and triple junctions in UO2 nuclear fuel

A phase-field model of fission gas bubble evolution was developed and applied to gain an improved understanding of the microstructure-level processes leading to fission gas release from nuclear fuel, and to inform engineering-scale fission gas release models. The phase-field model accounts for multiple fuel grains and fission gas bubbles and tracks the local concentration of vacancies and gas atoms. The model was used to simulate the growth of grain boundary and triple junction bubbles in a hexagonal periodic 3D grain structure. The fractional coverage of the triple junctions and the number of fully saturated triple junctions (that is, triple junctions fully covered by the gas phase) was calculated, and correlations were developed between these quantities and the grain boundary coverage. The effects of initial triple junction bubble density, vacancy source strength, and bubble semi-dihedral angle on triple junction coverage and saturation were evaluated. High initial triple junction coverage and high bubble semi-dihedral angle can lead to triple junction saturation well before grain boundary percolation. The implications of these findings for engineering-scale fuel performance modeling are discussed.

Kriging-based inverse uncertainty quantification of nuclear fuel performance code BISON fission gas release model using time series measurement data

In nuclear reactor fuel performance simulation, fission gas release (FGR) and swelling involve treatment of several complicated and interrelated physical processes, which inevitably depend on uncertain input parameters. However, the uncertainties associated with these input parameters are only known by “expert judgment”. In this paper, inverse Uncertainty Quantification (UQ) under the Bayesian framework is applied to BISON code FGR model based on Risø-AN3 time series experimental data. Inverse UQ seeks statistical descriptions of the uncertain input parameters that are consistent with the available measurement data. It always captures the uncertainties in its estimates rather than merely determining the best-fit values. Kriging metamodel is applied to greatly reduce the computational cost during Markov Chain Monte Carlo sampling.We performed a dimension reduction for the FGR time series data using Principal Component Analysis. We also projected the original FGR time series measurement data onto the PC subspace as “transformed experiment data”. A forward uncertainty propagation based on the posterior distributions shows that the agreement between BISON simulation and Risø-AN3 time series measurement data is greatly improved. The posterior distributions for the uncertain input factors can be used to replace the expert specifications for future uncertainty/sensitivity analysis.

Development of the MFPR model for fission gas release in irradiated UO2under transient conditions

The fission gas release microscopic model of the mechanistic code MFPR is further developed for modelling of enhanced release from irradiated UO2 fuel under transient conditions of the power ramp tests, along with the microstructure evolution characterised by the formation of a new population of large intragranular bubbles with a rather wide size distribution (from 30 to 500 nm), observed in transient-tested UO2 fuel samples. Implementation of the additional microscopic mechanisms results in a notable improvement of the code predictions (in comparison with the previous code version) for the fractional gas release in the Riso ramp tests with three different hold times of 3, 40 and 62 h at the terminal linear power of ≈40 kW/m.