MOOSE NTP modeling: current status, needs and challenges
Presentation given at Georgia Tech
33 ADVANCED PROPULSION SYSTEMS↗
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Impact of temperature variations on BISON predictions of Ag release in AGR-1 and AGR-2 experiments
Understanding and quantifying the release of fission products like silver (Ag) from TRistructural ISOtropic (TRISO) fuel particles is important to assess the safe operation of advanced high temperature reactors. Although the silicon carbide (SiC) layer of TRISO particles is effective as the main fission product barrier, Ag can be released from intact TRISO particles. A mechanistic model for the effective Ag diffusivity, D eff , was previously developed as a function of temperature and microstructure variables informed by atomistic modeling of Ag diffusivity on the mesoscale. Here in this study, we use this model to explore how experimental temperature uncertainties impact the overall predicted Ag release. This analysis shows that temperature uncertainties have a significant impact on the overall Ag release predictions. Furthermore, we show that the time average volume average temperature (TAVA) temperature is not an appropriate proxy for temperature histories to predict fission product release. We attribute this to the Arrhenius dependence of Ag diffusivity with respect to temperature. The detailed temperature histories, therefore, provide the most accurate results are are of most importance for modeling efforts. Overall, this work shows the importance of considering the experimental uncertainty of the temperature on computational predictions of fission product transport and release and the need for more accurate temperature histories from future experiments.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
BISON Fuel Performance Analysis of FeCrAl cladding with updated properties
In order to improve the accident tolerance of light water reactor (LWR) fuel, alternative cladding materials have been proposed to replace zirconium (Zr)-based alloys. Of these materials, there is a particular focus on iron-chromium-aluminum (FeCrAl) alloys due to much slower oxidation kinetics in high-temperature steam than Zr-alloys. This should decrease the energy release due to oxidation and allow the cladding to remain integral longer in the presence of high temperature steam, making accident mitigation more likely. As a continuation of the development for these alloys, suitability for normal operation must also be demonstrated. This research is focused on modeling the integral thermo-mechanical performance of FeCrAl-cladded fuel during normal reactor operation. Preliminary analysis has been performed to assess FeCrAl alloys (namely Alkrothal 720 and APMT) as a suitable fuel cladding replacement for Zr-alloys, using the MOOSE-based, finite-element fuel performance code BISON and the best available thermal-mechanical and irradiation-induced constitutive properties. These simulations identify the effects of the mechanical-stress and irradiation response of FeCrAl, and provide a comparison with Zr-alloys. In comparing these clad materials, fuel rods have been simulated for normal reactor operation and simple steady-state operation. Normal reactor operating conditions target the cladding performance over the rod lifetime (~4 cycles) for the highest-power rod in the highest-power fuel assembly under reactor power maneuvering. The power histories and axial temperature profiles input into BISON were generated from a neutronics study on full-core reactivity equivalence for FeCrAl using the 3D full core simulator NESTLE. Evolution of the FeCrAl cladding behavior over time is evaluated by using steady-state operating conditions such as a simple axial power profile, a constant cladding surface temperature, and a constant fuel power history. The fuel rod designs and operating conditions used are based off the Peach Bottom BWR and design consideration was given to minimize the neutronic penalty of the FeCrAl cladding by changing fuel enrichment and cladding thickness. As this study progressed, systematic parametric analysis of the fuel and cladding creep responses were also performed.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
BISON Robustness and Performance Improvements
BISON is a modern finite-element based nuclear fuel performance code that has been under development at the Idaho National Laboratory (USA) since 2009 [1]. The code is applicable to both steady and transient fuel behavior and can be used to analyze 1D (spherically symmetric), 2D (axisymmetric and generalized plane strain) or 3D geometries. BISON is the fuel performance code used within CASL for LWR fuel under both normal operating and accident conditions. BISON is built using the INL Multiphysics ObjectOriented Simulation Environment, or MOOSE [2, 3]. MOOSE is a massively parallel, finite element-based framework to solve systems of coupled non-linear partial differential equations using the Jacobian-Free Newton Krylov (JFNK) method [4]. This enables investigation of computationally large problems, for example a full stack of discrete pellets in a LWR fuel rod, or every rod in a full reactor core. MOOSE supports the use of complex two and three-dimensional meshes and uses implicit time integration, important for the widely varied time scale in nuclear fuel simulation. An object-oriented architecture is employed which greatly minimizes the programming effort required to add new material and behavioral models. The flexibility of the implicit and fully coupled multiphysics approach comes with a need for constructing suitable approximations for the Jacobian matrix of the coupled system used for either preconditioning a Krylov solve or in a direct Newton solve. Preconditioning options for Bison problems need to be revisited with new preconditioning methods becoming available.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Preliminary modeling of accident tolerant fuel concepts under accident conditions
In this paper, we present simulations of two proposed accident tolerant fuel systems for light-water reactors: U{sub 3}Si{sub 2} fuel with Zircaloy-4 cladding, and UO{sub 2} fuel with FeCrAl cladding. The simulations investigate the fuel performance response of the proposed ATF systems under Loss of Coolant (LOCA) and Station Blackout (SBO) conditions using the BISON code. It was observed that U{sub 3}Si{sub 2}/Zircaloy4 and UO{sub 2}/FeCrAl have superior performance during a 90 s LOCA than traditional UO{sub 2}/Zircaloy4 fuel rods. This is based upon reduced fuel centerline temperatures and lower cladding strains experienced during the LOCA. For the postulated SBO analyzed in this work, it was found that the centerline temperature and cladding strains remained well below any values of concern. This is attributed to the shortness of the rod studied and how the coolant channel model treats the inlet as the bottom of the rod. An analysis of a full-length fuel rod is required to obtain improved qualitative comparisons between the 3 systems under SBO conditions. Sensitivity analyses have been performed using Sandia National Laboratories' DAKOTA software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). Early results indicate that each concept has significant advantages as well as areas of concern. Further work is required prior to formulating the proposition report for the Advanced Fuels Campaign.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Research in Support of TREAT Kinetics Calculations Using Rattlesnake/BISON Coupling Within MAMMOTH
This paper summarizes key aspects of research in evaluation of modeling needs for TREAT transient simulation. Using measured data from transient experiments performed circa 1959 in a small, test core used for startup testing, MAMMOTH multi-physics simulations are per-formed to simulate the core transient behavior. MAMMOTH uses internal libraries from Rattlesnake for time-dependent transport calculations, BISON for fuel performance calcula-tions and the MOOSE multi-physics framework to perform strongly coupled calculations to properly characterize the three-dimensional temperature-driven kinetics performance of the core during rapid power excursions. Early calculations were able to match the general shape of measured transient data but under-predicted peak power. Closer examination of the phys-ics of the core and uncertainty associated with rod positioning led to calculations in which the best estimate of the actual measured reactor period was modeled. This correction resulted in outstanding agreement between measurements and coupled multi-physics simulation results.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Advanced multiphysics coupling for LWR fuel performance analysis
Even the most basic nuclear fuel analysis is a multiphysics undertaking, as a credible simulation must consider at a minimum coupled heat conduction and mechanical deformation. The need for more realistic fuel modeling under a variety of conditions invariably leads to a desire to include coupling between a more complete set of the physical phenomena influencing fuel behavior, including neutronics, thermal hydraulics, and mechanisms occurring at lower length scales. This paper covers current efforts toward coupled multiphysics LWR fuel modeling in three main areas. The first area covered in this paper concerns thermomechanical coupling. The interaction of these two physics, particularly related to the feedback effect associated with heat transfer and mechanical contact at the fuel/clad gap, provides numerous computational challenges. An outline is provided of an effective approach used to manage the nonlinearities associated with an evolving gap in BISON, a nuclear fuel performance application. A second type of multiphysics coupling described here is that of coupling neutronics with thermomechanical LWR fuel performance. DeCART, a high-fidelity core analysis program based on the method of characteristics, has been coupled to BISON. DeCART provides sub-pin level resolution of the multigroup neutron flux, with resonance treatment, during a depletion or a fast transient simulation. Two-way coupling between these codes was achieved by mapping fission rate density and fast neutron flux fields from DeCART to BISON and the temperature field from BISON to DeCART while employing a Picard iterative algorithm. Finally, the need for multiscale coupling is considered. Fission gas production and evolution significantly impact fuel performance by causing swelling, a reduction in the thermal conductivity, and fission gas release. The mechanisms involved occur at the atomistic and grain scale and are therefore not the domain of a fuel performance code. However, it is possible to use lower length scale models such as those used in the mesoscale MARMOT code to compute average properties, e.g. swelling or thermal conductivity. These may then be used by an engineering-scale model. Examples of this type of multiscale, multiphysics modeling are shown.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Sensitivity analysis, surrogate modeling, and optimization of pebble-bed reactors considering normal and accident conditions
This research provides a valuable tool that streamlines the optimization process while significantly increasing its accuracy. This study creates a robust framework for reactor design optimization by incorporating comprehensive modeling using the Comprehensive Reactor Analysis Bundle, or BlueCRAB, within the Multiphysics Object-Oriented Simulation Environment (MOOSE). BlueCRAB is the United States Nuclear Regulatory Commission's code suite for non-light water reactor analysis and includes the Griffin, Pronghorn, and Bison applications. This not only improves the efficiency of the optimization process but also enhances the reliability of the results. Such a tool is essential for advancing the state-of-the-art in pebble-bed reactor technology and is critical for achieving the goals of Generation IV reactors, which aim for safe, sustainable, and economically viable nuclear energy solutions. This work presents and applies this workflow on pebble-bed reactors while considering both normal and off-normal conditions. A representative gas-cooled pebble-bed reactor at equilibrium core conditions serves as the nominal design specification for normal operation and is based on previous research. The depressurized loss-of-forced-cooling accident is deployed for off-normal conditions in this work. After defining design-related parameters and quantities of interest regarding reactor safety and performance, this multiphysics model is sampled using the MOOSE stochastic tools module. The result is a comprehensive dataset of configurations, enabling sensitivity analysis and the generation of surrogate models. Subsequently, the dataset and surrogate models are employed in two optimization studies aimed at maximizing fuel utilization and economic profit while adhering to safety and operational constraints. Performing the optimization process with fuel utilization as the metric leads to an improvement of approximately 10%, compared to engineering-judgment-based nominal conditions. The optimization on economic profit leads to an estimated increase of ~300 million USD over the lifetime of the reactor.
97 MATHEMATICS AND COMPUTING↗
Thermal-Hydraulic Performance of the TREAT Multi-SERTTA for Reactivity Initiated Accident Experiments
Experimental testing in the Multi-Static Environment Rodlet Transient Test Apparatus (SERTTA) will lead the rebirth of transient fuel testing in the United States as part of the Accident Tolerant Fuels (ATF) progam. The Multi-SERTTA is comprised of four isolated pressurized environments capable of a wide variety of working fluids and thermal conditions. Ultimately, the TREAT reactor as well as the Multi-SERTTA test vehicle serve the purpose of providing desired thermal-hydraulic boundary conditions to the test specimen. The initial ATF testing in TREAT will focus on reactivity insertion accident (RIA) events using both gas and water environments including typical PWR operating pressures and temperatures. For the water test environment, a test configuration is envisioned using the expansion tank as part of the gas-filled expansion volume seen by the test to provide additional pressure relief. The heat transfer conditions during the high energy power pulses of RIA events remains a subject of large uncertainty and great importance for fuel performance predictions. To support transient experiments, the Multi-SERTTA vehicle has been modeled using RELAP5 with a baseline test specimen composed of UO2 fuel in zircaloy cladding. The modeling results show the influence of the designs of the specimen, vehicle, and transient power pulses. The primary purpose of this work is to provide input and boundary conditions to fuel performance code BISON. Therefore, studies of parameters having influence on specimen performance during RIA transients are presented including cladding oxidation, power pulse magnitude and width, cladding-to-coolant heat fluxes, fuel-to-cladding gap, transient boiling effects (modified CHF values), etc. The results show the great flexibility and capacity of the TREAT Multi-SERTTA test vehicle to provide testing under a wide range of prototypic thermal-hydraulic conditions as never done before.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
M3 milestone: Advanced contact 2021
This report describes advanced contact approaches and implementations in FY21 for use in emerging reactors with geometries that differ from LWR/EBR-II cylindrical fuel. Contact remains a challenging and crucially important component of our modeling toolkit. While significant improvements have been made through mortar and automatic differentiation, we explored advanced approaches (i.e., dual mortar) that yields better conditioned systems and will improve the robustness and performance of BISON contact problems. This effort is relevant to emerging reactor types, such as micro reactors and the corresponding complex and numerous contact surfaces. In this report, improvements of the mortar-based contact formulation are described. Benchmark problems and BISON assessment cases are included to demonstrate the improvements and advantages that are brought by the recent capability. Remaining issues are discussed and corresponding development plans are outlined.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Coupling of Thermochemistry Solver THERMOCHIMICA with MOOSE/BISON
THERMOCHIMICA is a computational library for chemical thermodynamics. It determines a unique combination of phases and their compositions for a prescribed chemical composition, temperature and pressure. The solver and the underlying thermodynamic models can be used to estimate the chemical state and various constitutive and transport properties necessary for nuclear fuel performance modeling. Our NEAMS research project focuses on the development of nuclear fuel thermodynamics models and computational thermodynamics equilibrium solvers, and their implementation into Idaho National Laboratory (INL) nuclear fuel performance simulation system MOOSE/BISON. This report documents the software integration of THERMOCHIMICA with MOOSE/BISON using the current INL development framework GitLab. The THERMOCHIMICA implementation is demonstrated on modeling oxygen related phenomena in irradiated fuel, such as calculation of oxygen to metal ratio in the fluorite phase, oxygen partial pressure, oxygen chemical potential and oxygen transport. Experimental measurements from the open literature were used to validate the implemented models and illustrate functionality of the developed thermodynamics module. The calculations are based on chemical element inventory provided by neutronics, isotopic depletion, transmutation and decay calculations in the SCALE system. The burnup calculation in BISON can also be used for elemental composition although it provides only a limited set. Coupling of THERMOCHIMICA with other codes that can provide chemical composition of irradiated fuel for BISON simulations is currently being explored.
37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
Development of a creep model informed by lower-length scale simulations to simulate creep in doped UO 2
Using molecular dynamics, we predict information at the atomistic scale used to develop a mechanistic UO 2 creep model for use in higher length-scale fuel performance codes. The ultimate objective of the model is to better describe the grain size dependence and therefore impact of doping on creep rates in UO 2 . In a previous NEAMS milestone, we found that Nabarro-Herring (bulk diffusional) creep was too low to capture the experimentally observed creep rates in standard UO 2 . Moreover, in that milestone, other mechanisms were explored, such as Coble (grain boundary) creep and dislocation climb, with each mechanism exhibiting different grain size dependencies. Again, these were orders of magnitude too low to describe the experimental creep rates. In this work, we address the previous assumptions made for the Coble creep mechanism by investigating the diffusivity of various defects at grain boundaries in UO 2 and, critically, to determine if enhanced grain boundary diffusivity allows the model to better reproduce experimental results. The diffusivity as a function of temperature for different concentrations of uranium vacancies and interstitials for bulk UO 2 have also been examined using cluster dynamics. Furthermore, using a concentration dependent segregation model, the concentration of defects at the grain boundary were predicted. This atomistic data was then input into the various creep mechanisms and the creep rates compared to the empirical MATPRO correlation (used in BISON) and experiment.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Advanced Multiphysics Code Coupling for Cladding Surface Thermocouples During Two-Phase Heat Transfer from Nuclear Fuel
Transient testing of nuclear fuel involves the evaluation of fuel performance under off-normal and accident conditions and is essential for proving design performance. Instrumentation included in such experiments commonly includes thermocouples attached to the outer surface of the cladding to provide temperature measurements throughout the transient. However, the presence of thermocouples on the cladding surface can alter the local heat transfer characteristics with the surrounding coolant. These localized effects can influence the temperature of the nearby cladding surface and introduce uncertainties in interpreting the thermocouple data. Understanding the impact of thermocouples attached to the outer surface of the cladding is crucial for accurate data interpretation as well as its effect on the thermomechanical behavior of the cladding. This paper presents a novel methodology for simulating the impact of outer cladding thermocouples during transient testing of nuclear fuels. The simulation framework leverages the thermal-hydraulic capabilities of RELAP5-3D coupled to the BISON fuel performance code through the RELAPCouplingApp interface. The methodology is compared against Accident Tolerant Fuel Reactivity Initiated Accident-1-E experiment performed at Idaho National Laboratory. The results reveal approximately 100°C difference between thermocouple-altered temperature and virgin cladding surface. The model overpredicts surface rewet time due to conservative correlations.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
PRELIMINARY EVALUATION OF FeCrAl CLADDING AND U-Si FUEL FOR ACCIDENT TOLERANT FUEL CONCEPTS
Since the accident at the Fukushima Daiichi Nuclear Power Station, enhancing the accident tolerance of light water reactors (LWRs) has become an important research topic. In particular, the community is actively developing enhanced fuels and cladding for LWRs to improve safety in the event of accidents in the reactor or spent fuel pools. Fuels with enhanced accident tolerance are those that, in comparison with the standard UO2-zirconium alloy system, can tolerate loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations and operational transients. This paper presents early work in developing thermal and mechanical models for two materials that may have promise: U-Si for fuel, and FeCrAl for cladding. These materials would not necessarily be used together in the same fuel system, but individually have promising characteristics. BISON, the finite element-based fuel performance code in development at Idaho National Laboratory, was used to compare results from normal operation conditions with Zr-4/UO2 behavior. In addition, sensitivity studies are presented for evaluating the relative importance of material parameters such as ductility and thermal conductivity in FeCrAl and U-Si in order to provide guidance on future experiments for these materials.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Coupling RELAP5-3D to BISON
This report details two approaches for coupling the BISON nuclear fuel performance code with RELAP5-3D. Both approaches are shown to work well. The first is a Python-based approach, and the second combines the two applications into one executable with data passed in memory from one library to the other. This second coupling approach forms the basis for analysis of complex loss of coolant scenarios and enables future calculations of modeling uncertainties.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Report on Assessment of Advanced Reactor Fuel Performance
Three activities have been performed in FY 2019 to support the Advanced Fuel Cycle Campaign, which include assessment of annular fuel performance in an ultra-high burnup sodium-cooled reactor (SFR), modeling of annular fuel behavior during a transient scenario, and feasibility of a sustainable closed fuel cycle utilizing melt-refining process. Annular fuel performance parameters, including temperature profile, swelling strain, and cumulative damage fraction (CDF), in a 3000 MWt Sustainable Sodium-cooled Fast Reactor (SSFR) were assessed using the BISON fuel performance code. The SSFR is once-through ultra-long-life core based on a 34-batch fuel management scheme with 1.5-year cycle length. The fuel residence time in the SSFR is 51 years and the average discharged burnup is about 30%. Annular fuel has potential to irradiate for ultra-high burnup without major fuel-cladding mechanical interaction. This feature is achieved by using a low smeared density (~55%). In the SSFR design, however, 75% smeared density and a thick fuel pin were preferred to maintain criticality for 51 years. Due to the high smeared density and thick fuel pin configuration, the annular fuel was predicted to suffer premature cladding failure before reaching the discharge burnup of the SSFR. The current fuel creep model predicts a prominent creep strain due to gravity over the half-century irradiation, which may significantly affect the evolution of CDF. Thus, in order to use the annular fuel in an ultra-high burnup core such as the SSFR, a low smeared density and thin annular fuel is strongly recommended.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
A layered 2D computational framework: Theory and applications to nuclear fuel behavior
Nuclear fuel performance computer codes have been developed over the last 50 years to analyze fuel behavior under various operating conditions. Traditionally, these codes used quasi-two-dimensional (also commonly known as 1.5D) representations of the fuel rod, which model the rod using a set of one-dimensional axisymmetric models that represent the behavior at specific axial positions on the rod. Modern fuel performance codes have the ability to investigate full three-dimensional (3D) effects and couple to other physics-based codes for true multiphysics simulations. However, with increasing complexity comes increasing computational costs. Many phenomena of interest involve azimuthally-varying behavior that cannot be represented using the aforementioned quasi-two-dimensional approach, but do not require the use of a full 3D model. To efficiently address these problems, there is a need for a computational framework that provides a compromise between the quasi-two-dimensional and full 3D models. In this paper, we present a new quasi-three-dimensional approach that represents the fuel as a set of 2D planar models that represent the behavior of the fuel cross-section at various axial positions. Presented here are the theory behind the methodology, test cases to illustrate proper implementation, and practical applications of its use in the BISON fuel performance code for a variety of cases in nuclear fuel analysis, such as fuel fracture, axial fuel relocation, and cladding distension and oxidation.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Thermo-mechanical Analysis of SiC/SiC Composite Cladding with BISON including Fuel Creep
This report presents the results of a parametric thermo-mechanical parametric analysis of SiC/SiC cladding performed using the BISON fuel performance code. In this analysis different steady state linear heat rates (LHR) (18kW/m, 22kW/m, 26kW/m and 30kW/m) and initial fuel rod-gap thicknesses (50μm, 65μm, 80μm and 95μm) were considered along with three different model implementation of fuel creep. These include 1) no fuel creep 2) fuel creep based on the Hagrman model and 3) fuel creep based on a modified model. The fuel rod geometries used in this analysis are representative of a generic 17 x 17 fuel rod assembly used in a Westinghouse PWR.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗