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BISON is a nuclear fuel performance application built using the Multiphysics Object-Oriented Simulation Environment (MOOSE) finite element library. One of its major goals is to have a great amount ...

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity multiphysics simulations of light water nuclear reactors. To accomplish this, CASL is developing the Virtual Environment for Reactor Applications (VERA), which is a suite of code packages for thermal hydraulics, neutron transport, fuel performance, and coolant chemistry. As VERA continues to grow and expand, there has been an increased focus on incorporating fuel performance analysis methods. One of the primary goals of CASL is to estimate local cladding failure probability through pellet-clad interaction, which consists of both pellet-clad mechanical interaction (PCMI) and stress corrosion cracking. Estimating clad failure is important to preventing release of fission products to the primary system and accurate estimates could prove useful in establishing less conservative power ramp rates or when considering load-follow operations. While this capability is being pursued through several different approaches, the procedure presented in this article focuses on running independent fuel performance calculations with BISON using a file-based one-way coupling based on multicycle output data from high fidelity, pin-resolved coupled neutron transport–thermal hydraulics simulations. This type of approach is consistent with traditional fuel performance analysis methods, which are typically separate from core simulation analyses. A more tightly coupled approach is currently being developed, which is the ultimate target application in CASL. Recent work simulating 12 cycles of Watts Bar Unit 1 with VERA core simulator are capitalized upon, and quarter-core BISON results for parameters of interest to PCMI (maximum centerline fuel temperature, maximum clad hoop stress, and minimum gap size) are presented for Cycles 1–3. Based on these results, this capability demonstrates its value and how it could be used as a screening tool for gathering insight into PCMI, singling out limiting rods for further, more detailed analysis.

Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2’s fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety of LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.

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

U3Si2 is an advanced fuel candidate due to its relatively high fissile density and attractive thermal properties. Compared to standard UO2 fuel, there are significant data gaps for the thermophysical and thermomechanical properties of U3Si2. Point defect concentrations and mobilities under irradiation govern a number of important fuel performance properties, such as creep and fission gas release. In this work, we utilized density functional theory (DFT) data to inform a cluster dynamics framework to predict point defect concentrations in U3Si2 under irradiation. Molecular dynamics (MD) simulations were used to examine the contribution of atomic mixing during ballistic cascades to diffusion, as well as the diffusivity of U and Si at grain boundaries. These atomic scale models for diffusivity were then used to inform a creep model based on bulk (Nabarro-Herring) and grain boundary (Coble) diffusional creep, and climb-limited dislocation creep. The model compares well against available experimental data and has been implemented in the BISON fuel performance code. A demonstration case using simple power profiles has been carried out, showing that negligible creep occurs due to the low temperatures experienced by U3Si2 in-reactor, a consequence of its high thermal conductivity.

Modeling of tristructural isotropic (TRISO)-coated particle fuel is being refined in the fuel performance code BISON. New developments include the implementation of an updated set of material properties, TRISO failure mechanisms, fission product diffusion parameters, and the design of a Monte Carlo scheme that allows BISON to calculate the probability of fuel failure within a population of TRISO particles and the subsequent fractional release of key fission products.

Rate theory simulations of fission gas behavior in U 3 Si 2 are reported for light water reactor (LWR) steady-state operation scenarios. A model of U 3 Si 2 was developed and implemented into the GRASS-SST code based on available research reactor post-irradiation examination (PIE) data, and density functional theory (DFT) calculations of key material properties. The reliability of the model was examined by performing sensitivity analysis of the key parameters. Simplified peripheral models were also introduced to capture the fuel-cladding interaction. The simulations identified three regimes of U 3 Si 2 swelling behavior between 390 K and 1190 K. Under typical steady-state LWR operating conditions where U 3 Si 2 temperature is expected to be below 1000 K, intragranular bubbles are dominant and fission gas is retained in those bubbles. The consequent gaseous swelling is low and associated degradation in the fuel thermal conductivity is also limited. Those predictions of U 3 Si 2 performance during steady-state operations in LWRs suggest that this fuel material is an appropriate LWR candidate fuel material. Fission gas behavior models established based on this work are being coupled to the thermo-mechanical simulation of the fuel behavior using the BISON fuel performance multi-dimensional finite element code.

In this work, we present a model of fission gas behavior in U3Si2 under light water reactor (LWR) conditions for application in engineering fuel performance codes. The model includes components for intra-granular and inter-granular behavior of fission gases. The intra-granular component is based on cluster dynamics and computes the evolution of intra-granular fission gas bubbles and swelling coupled to gas diffusion to grain boundaries. The inter-granular component describes the evolution of grain-boundary fission gas bubbles coupled to fission gas release. Given the lack of experimental data for U3Si2 under LWR conditions, the model is informed with parameters calculated via atomistic simulations. In particular, we derive fission gas diffusivities through density functional theory calculations, and the re-solution rate of fission gas atoms from intra-granular bubbles through binary collision approximation calculations. The developed model is applied to the simulation of an experiment for U3Si2 irradiated under LWR conditions available from the literature. Results point out a credible representation of fission gas swelling and release in U3Si2. Finally, we perform a sensitivity analysis for the various model parameters. Based on the sensitivity analysis, indications are derived that can help in addressing future research on the characterization of the physical parameters relative to fission gas behavior in U3Si2. The developed model is intended to provide a suitable infrastructure for the engineering scale calculation of fission gas behavior in U3Si2 that exploits a multiscale approach to fill the experimental data gap and can be progressively improved as new lower-length scale calculations and validation data become available.

The diffusion of fission products in intact TRISO particles depends on particle geometry, fission product source rates, time, temperature, and temperature-dependent diffusion coefficients. Simulating this diffusion process requires models for source rates and diffusion coefficients, plus a computation of the temperature field if not prescribed. In addition, the simulation quality depends on the discretization of the geometry, appropriate time stepping, and the accuracy of the solution method. In this paper, we explore the simulation of fission product diffusion in TRISO fuel particles using the finite element method via the fuel performance code BISON. Recent material model development has occurred in BISON for each material present in tri-structural isotropic (TRISO) fuel particles: the buffer, inner pyrolytic carbon, silicon carbide, and outer pyrolytic carbon layers, as well as the fuel kernel. Also, new mesh generation and fission product release fraction capabilities have been added. Diffusion capabilities are shown to converge to the correct solution via formal verification tests. A large number of code benchmarking problems are also given, with good results, showing that BISON’s computed release fractions closely match those of other software tools. Finally, a significant validation effort is detailed in which fission product release, measured as part of the AGR-1 capsule experiments, is compared to BISON outputs. BISON outputs compare favorably to the experimental data and to PARFUME results.

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.

In the first part of this work, we discussed BISON fuel performance code developments for Zircaloy cladding behavior under loss-of-coolant accident (LOCA) conditions and validation to separate-effects experiments. In this paper, we present modeling developments for UO2 fuel behavior during LOCAs and validation to integral experiments. Code developments are related in particular to modeling axial relocation of fuel fragments during cladding ballooning. Code validation is performed against three integral fuel rod LOCA tests from the Halden IFA-650 series, covering fresh and pre-irradiated fuel rods with and without significant axial fuel relocation observed. Calculated results are systematically compared to experimental data of fuel rod inner pressure evolution during the LOCA transients, time to cladding burst and post-test cladding diameter profile. Comparisons in terms of rod pressure evolution and time to burst appear adequate. Also, the ability to reproduce axial relocation of fuel fragments is demonstrated. Calculated cladding diameter profiles are shown to be considerably sensitive to the selection of the burst failure criterion, highlighting the difficulty to accurately evaluate cladding strains at burst during postulated LOCA scenarios with fuel performance codes.

U3Si2 is a candidate accident tolerant fuel (ATF) replacement for UO2. U3Si2’s high uranium density and high thermal conductivity are favorable properties in steady-state and accident conditions. Low power performance of this U3Si2-SiC concept fuel is compared to that of UO2-Zr4 fuels by implementing models that describe the properties of U3Si2 and SiC-SiC into Idaho National Laboratory's (INL) fuel performance code, BISON. Included in these material models is a thermal creep model for U3Si2 based on compressive creep data. The simulated results are in keeping with community knowledge that the U3Si2-SiC concept fuel may serve as a replacement for UO2-Zr4 fuels during steady-state operation, provided the mSiC layer remains under compression. Through a moderate power history and three 24-month fuel cycles, the mSiC layer remains under compressive stress through a burnup of 80 MWd/kgU. During low power operation, failure of the mSiC layer generally occurs prior to significant thermal creep in U3Si2. Generally, U3Si2 creep is temperature sensitive and of little importance at the temperatures and stresses simulated during steady operation and during fuel-to-cladding contact. A parameter variation study including 11,520 individual simulations with variations in nominal fuel thermal creep rate, cladding thermal conductivity, cladding irradiation creep and swelling, cladding gap size, and cladding thickness demonstrated that research priorities for this ATF should revolve around reducing cladding thickness as a means to minimize cladding failure. Generally, despite advances in SiC-SiC compliance, the brittle nature of mSiC excludes U3Si2-SiC for use where fuel cladding contact may occur.

U3Si2 is a potential accident-tolerant fuel that shows promise due to its high thermal conductivity and higher uranium density relative to UO2. However, its swelling and fission gas release behavior in light water reactor (LWR) conditions is relatively unknown. To provide mechanistic insight and determine parameters for engineering-scale fuel performance modeling of pellet-form U3Si2, phase-field simulations of the growth, interconnection, and venting of intergranular fission gas bubbles were performed. The fractional coverage of the grain boundary and the fraction of bubble area that is vented were calculated as a function of time. From the simulation data, the fractional grain boundary coverage at saturation, an important parameter needed in engineering-scale modeling of swelling and fission gas release, was determined. Multiple simulations were run to determine the uncertainty in the calculated value. The effect of model assumptions and input parameters that are not well known was evaluated. Simulation results are compared to related theoretical and computational work. Based on the simulation results, a value of 0.60 for the fractional grain boundary coverage at saturation is recommended for U3Si2 fuel.

BISON is a modern finite element-based nuclear fuel performance code that has been under development at Idaho National Laboratory (INL) since 2009. The code is applicable to both steady and transient fuel behavior and has been used to analyze a variety of fuel forms in 1D spherical, 2D axisymmetric, or 3D geometries. Code validation is underway and is the subject of this study. A brief overview of BISON's computational framework, governing equations, and general material and behavioral models is provided. BISON code and solution verification procedures are described, followed by a summary of the experimental data used to date for validation of Light Water Reactor (LWR) fuel. Validation comparisons focus on fuel centerline temperature, fission gas release, and rod diameter both before and following fuel-clad mechanical contact. Comparisons for 35 LWR rods are consolidated to provide an overall view of how the code is predicting physical behavior, with a few select validation cases discussed in greater detail. Results demonstrate that (1) fuel centerline temperature comparisons through all phases of fuel life are very reasonable with deviations between predictions and experimental data within ±10% for early life through high burnup fuel and only slightly out of these bounds for power ramp experiments, (2) accuracy in predicting fission gas release appears to be consistent with state-of-the-art modeling and with the involved uncertainties and (3) comparison of rod diameter results indicates a tendency to overpredict clad diameter reduction early in life, when clad creepdown dominates, and more significantly overpredict the diameter increase late in life, when fuel expansion controls the mechanical response. Initial rod diameter comparisons are not satisfactory and have led to consideration of additional separate effects experiments to better understand and predict clad and fuel mechanical behavior. Results from this study are being used to define priorities for ongoing code development and validation activities.

Horizontally oriented fuel bundles, such as those in CANada Deuterium Uranium (CANDU) reactors present unique modeling challenges. After long irradiation times or during severe transients the fuel elements can laterally deform out of plane due to processes known as bow and sag. Bowing is a thermally driven process that causes the fuel elements to laterally deform when a temperature gradient develops across the diameter of the element. Sagging is a coupled mechanical and thermal process caused by deformation of the fuel pin due to creep mechanisms of the sheathing after long irradiation times and or high temperatures. These out-of-plane deformations can lead to reduced coolant flow and a reduction in coolability of the fuel bundle. In extreme cases element-to-element or element-to-pressure tube contact could occur leading to reduced coolant flow in the subchannels or pressure tube rupture leading to a loss of coolant accident. This paper evaluates the capability of the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework developed at the Idaho National Laboratory to model these deformation mechanisms. The material model capabilities of MOOSE and its ability to simulate contact are also investigated.

Iron-chromium-aluminum (FeCrAl) alloys are candidates to be used as nuclear fuel cladding for increased accident tolerance. An analysis of the response of FeCrAl under normal operating and loss of coolant conditions has been performed using fuel performance modeling. In particular, recent information on FeCrAl material properties and phenomena from separate effects tests has been implemented in the BISON fuel performance code and analyses of integral fuel rod behavior with FeCrAl cladding have been performed. BISON simulations included both light water reactor normal operation and loss-of-coolant accidental transients. In order to model fuel rod behavior during accidents, a cladding failure criterion is desirable. For FeCrAl alloys, a failure criterion is developed using recent burst experiments under loss of coolant like conditions. The added material models are utilized to perform comparative studies with Zircaloy-4 under normal operating conditions and oxidizing and non-oxidizing out-of-pile loss of coolant conditions. The results indicate that for all conditions studied, FeCrAl behaves similarly to Zircaloy-4 with the exception of improved oxidation performance. Further experiments are required to confirm these observations.

The effect of various p-method finite element model parameters on the prediction of planar crack front geometry in three dimensional structures is evaluated. An automatic crack growth method was developed using the commercial software StressCheck coupled with Microsoft Excel. The geometry of the evolving crack front was predicted using iterations of three dimensional finite element models to determine the stress intensity values at discrete points along the crack front and incrementing the crack length locally using the Paris growth law. Modeling parameters such as the global mesh size, p-level, crack tip mesh size, and the extraction radius used for determining the stress intensity factors were all considered. The effect of these parameters on the predicted shape of a growing elliptical crack front was evaluated through a series of simulations of a corner crack emanating from a centered hole in a plate. A single model of a through crack with a straight crack front was also used to compare the effects of the model parameters on a different geometric configuration. This study has shown that the corner crack geometry and the growth series are more sensitive to the number of elements along the crack front and the extraction radius than the through thickness crack. In addition, convergence of the predicted crack front geometry does not guarantee convergence of the distribution of stress intensity factors along the crack front. It was also determined that for the models considered, using a smaller overall mesh size is a more effective way of decreasing discretization error than increasing the p-level.

This report summarizes a NEAMS (Nuclear Energy Advanced Modeling and Simulation) project focused on sensitivity analysis of a fuels performance benchmark problem. The benchmark problem was defined by the Uncertainty Analysis in Modeling working group of the Nuclear Science Committee, part of the Nuclear Energy Agency of the Organization for Economic Cooperation and Development (OECD ). The benchmark problem involv ed steady - state behavior of a fuel pin in a Pressurized Water Reactor (PWR). The problem was created in the BISON Fuels Performance code. Dakota was used to generate and analyze 300 samples of 17 input parameters defining core boundary conditions, manuf acturing tolerances , and fuel properties. There were 24 responses of interest, including fuel centerline temperatures at a variety of locations and burnup levels, fission gas released, axial elongation of the fuel pin, etc. Pearson and Spearman correlatio n coefficients and Sobol' variance - based indices were used to perform the sensitivity analysis. This report summarizes the process and presents results from this study.

This summary report contains an overview of work performed under the work package entitled “FY2014 NEAMS INL-Engineering Scale Fuel Performance & Interface with RPL Tools.” A first chapter identifies the specific FY-14 milestones, providing a basic description of the associated work and references to related detailed documentation. Where applicable, a representative technical result is provided. A second chapter summarizes substantial additional work including 1) efforts to improve numerical convergence and contact in BISON, 2) development of capability to simulate hydrogen behavior in Zircaloy cladding and 3) efforts to enhance collaborative work with the Halden Research Program. A final chapter briefly outlines planned future work.

The BISON code is the foundation for multiple fuel performance modeling efforts, and is cur- rently under heavy development. For a variety of fuel forms, the effects of heat conduction across a gap and mechanical contact between components of a fuel system are very significant. It is thus critical that BISON have robust capabilities for enforcement of thermal and mechanical contact. BISON’s solver robustness has generally been quite good before mechanical contact between the fuel and cladding occurs, but there have been significant challenges obtaining converged so- lutions once that contact occurs and the solver begins to enforce mechanical contact constraints. During the current year, significant development effort has been focused on the enforcement of mechanical contact to provide improved solution robustness. In addition to this work to improve mechanical contact robustness, an investigation into ques- tionable results attributable to thermal contact has been performed. This investigation found that the order of integration typically used on the surfaces involved in thermal contact was not suffi- ciently high. To address this problem, a new option was provided to permit the use of a different integration order for surfaces, and new usage recommendations were provided.

A three-dimensional finite element model is being developed for a quarter fuel element, which is equivalent to a full fuel element using symmetry. The model uses the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework developed at Idaho National Laboratory. The model facilitates an in-depth investigation into a variety of deformation phenomena for a horizontal nuclear fuel element including bowing, sagging, and stresses and strains. This paper presents a preliminary analysis of the local stresses and strains of the sheath (clad) at the pellet-to-pellet interfaces for low, normal and high linear powers. During irradiation the fuel pellets thermally expand and take on an hourglass shape. The hourglassing behaviour leads to higher local stresses and strains in the sheath at the locations of the pellet-to-pellet interfaces. The purpose of this work is to quantify these stresses and strains for varying linear powers, and to illustrate the effect that the material model chosen for the cladding has on the results. Preliminary results are presented for two sheath types: elastic, and elastic including diffusional creep. These models are benchmarked against a validated industry code called ELESTRES. The results indicate that the predicted sheath hoop strain is about half of what is determined by ELESTRES in both the elastic and elastic-creep cases. This highlights the requirement of a pellet cracking model in three-dimensional simulations. The elastic-creep model predicts less stress within the sheath than the elastic model as expected.