Machine learning-assisted surrogate construction for full-core fuel performance analysis
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Nuclear Science & Technology↗
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Enabling BWR fuel rod analysis in the BISON fuel performance code
Nuclear fuel vendors around the world are pursuing approaches to sustain the existing nuclear reactor fleet consisting primarily of pressurized-water reactors (PWRs) and boiling-water reactors (BWRs). To support the industry's efforts, advanced modeling and simulation tools need to be capable of analyzing both legacy reactor concepts. BWR fuel rods are significantly different than those used in PWRs, which can affect fuel performance analysis. BWR fuel rods include: (1) an extensive use of gadolinia dopant as a burnable absorber, (2) an axial variation in fuel enrichment and gadolinia content, (3) the inclusion of a liner on the inner cladding surface to mitigate the impact of pellet-clad mechanical interaction (which impacts hydrogen and hydride distribution), (4) a lower initial fill gas pressure, (5) bottom-entry control rods, and (6) a lower coolant pressure that results in the two-phase flow boiling phenomenon. Although the primary focus of BISON has been in the area of PWR and advanced reactor fuel analyses, this paper presents the developments in BISON to support BWR fuel performance analysis. An overview of the models that account for the effects of gadolinia is highlighted. Internal mesh generation capabilities to include a liner is presented. Normal operating and transient (reactivity insertion accident) demonstration problems are presented to illustrate the impact of gadolinia, the hydrogen and hydride evolution due to the presence of the liner, and BISON's ability to simulate axially varying enrichments and dopant concentration. Bottom-entry control effects are captured by the axial power peaking factors present in the demonstration cases. Comparisons to integral experiments from the Halden IFA-681 experiments are discussed as initial validation. Reasonable comparisons are obtained for fuel centerline temperature and rod internal pressure as a function of time. In conclusion, simulations of additional experiments containing Gd 2 O 3 -bearing fuel are necessary to completely validate the code for BWR applications.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
PCI analysis of a commercial PWR using BISON fuel performance code
Through the early 2000’s, nuclear fuel cladding failure resulting from pellet-cladding interaction in commercial pressurized water reactor fuel was managed successfully by incorporating power ramp restrictions. Within the past decade, however, cladding failures have again begun to appear, often correlated to missing pellet surface defects, and further aggravated by modern reactor operating strategies demanding higher fuel duty operations. In response, renewed efforts in fuel behavior modeling and simulation are being pursued to improve our understanding of the conditions that can lead to missing pellet surface related cladding failure. This paper describes a methodology used to model pellet-cladding interaction related failures observed in commercial pressurized water reactors using the BISON fuel performance code developed at Idaho National Laboratory. In order to accomplish this goal, a three-step process is followed: 1) A failure-stress threshold is developed by using BISON code to model ramp test with known pellet-clad interaction related failures, 2) BISON analysis is applied to a commercial pressurized water reactor cycle startup with one rod confirmed to contain a missing pellet surface defect, and 3) an missing pellet surface sensitivity study of a cycle startup of the same commercial pressurized water reactor which contained a failed fuel rod of un-identified type. The analysis results described in this paper serve to illustrate the BISON code capabilities for modeling pressurized water reactor fuel subjected to higher duty power events.
Nuclear Science & Technology↗
Impact of uranium oxide (UO 2 ) fuel with molybdenum (Mo) inserts on pressurized water reactor performance and safety
This work investigates nuclear reactor performance and safety characteristics of UO 2 with high thermal conductivity Mo insert structures by using multiphysics modeling techniques. Additionally, the purpose of this study is to use scoping analyses to quantify the impact of using Mo inserts from neutronic and heat transfer standpoints. Attention is given to reactor performance parameters, such as cycle length, maximum fuel temperature, temperature gradients in the fuel, and stored energy in the fuel. The finite-element code BISON and the Monte Carlo particle transport code Serpent were used to perform sensitivity analyses on the Mo insert geometry to optimize the insert design and inform larger scale modeling that required the homogenization of the UO 2 and Mo. Although BISON is often used as a fuel performance analysis tool, it is used in this context for heat transfer analysis only. Fuel performance optimization is outside the scope of the current study, but would be important for future work focused on this concept. The results showed that the insert had little impact on neutronic performance and that homogenizing the UO 2 and Mo was acceptable for reactor physics calculations. Reactivity temperature coefficients calculated using homogeneous UO 2 -Mo were shown to be relatively similar to UO 2 , but higher Mo content and 235 U enrichment can reduce the worth of soluble boron and control rods. The effect of insert geometry on heat transfer was much greater, and an approximately 15–20% difference in maximum fuel temperature was predicted between the best and worst performing heat transfer geometries. Furthermore, thermal conductivity calibration based on the finite element analysis results was performed to improve the accuracy of temperature predictions in reactor analysis models that homogenized the UO 2 -Mo fuel. Compared with UO 2 in a pressurized water reactor (PWR), the optimized UO 2 -Mo design increased the margin to fuel melt by 13–32% across the fuel cycle, but it requires the 235U enrichment to exceed 5% to match the cycle length of conventional UO 2 .
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Assessing pellet-clad interaction with VERA: WBN1, cycles 6-7
The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high fidelity simulations of nuclear reactors through the development of the Virtual Environment for Reactor Applications (VERA). One of the primary goals of CASL is to estimate local cladding failure through pellet-clad interaction (PCI). While this capability is being pursued through several different approaches, the procedure presented here focuses on running independent fuel performance calculations with the BISON fuel performance code being developed by Idaho National Laboratory using the multicycle output data from coupled neutron transport/thermal-hydraulics simulations generated by MPACT, CTF, and MAMBA. This type of approach is consistent with traditional fuel performance analysis methods, where the fuel performance analyses are typically separate from core simulation analyses. Recent work simulating 12 cycles of Watts Bar Unit 1 (WBN1) with VERA are capitalized upon, and stand-alone quarter-core BISON results can be obtained for several cycles, assessing parameters of interest to PCI (e.g., maximum centerline fuel temperature, maximum clad hoop stress, minimum gap thickness, and cumulative damage index). This is accomplished by using the coupled neutronics/thermal-hydraulics output power and temperature distributions as inputs for each BISON fuel rod case. Building on previous work which demonstrated this capability for Cycles 1 through 3, this paper presents the results for Cycles 6 and 7. Cycle 7 is of particular interest because significant CRUD-induced power shift was observed. While the current results do not immediately indicate that any rods are at an extreme risk of failure, further analysis on rods highlighted by this work may shed additional light on the likelihood of failure. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Steady-state fuel performance analyses for the preliminary fuel concept of general atomics fast modular reactor
Here this manuscript presents the fuel performance analysis results of the General Atomics Fast Modular Reactor (FMR) based on an axi-symmetric (2D-RZ) geometry. Three fuel performance model sets that fit the FMR fuel specifications best, i.e., a BISON baseline model set, a BISON diffusion enhancement model set, and a BISON-FASTGRASS model set, were identified and evaluated against a series of relevant experimental cases featuring high burnup and low irradiation temperature conditions. The three BISON-based model sets were then utilized to conduct a comprehensive fuel performance analysis of the FMR fuel under normal operation including the shutdown/restarting periods for refueling. The evaluation of the fuel performance parameters, represented by temperature, internal pressure, stress, and strain, shows that the FMR fuel maintains its thermal and mechanical integrity during normal operation. Technology gaps and limitations are also discussed to guide future efforts for extending the performance analysis to transient scenarios as well as improving the fuel performance evaluation through experiments.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Kriging-Based Inverse Uncertainty Quantification of BISON Fission Gas Release Model
Nuclear reactor fuel performance analysis studies the thermo-mechanical behavior of fuel rods and verify their compliance with safety criteria under both normal operation and accidental conditions. One of the major concerns in fuel performance analysis is the behavior of the fission gases xenon and krypton in uranium dioxide fuel, which significantly affect the thermo-mechanical performance of the nuclear fuel rods employed in current LWRs. Previously, the modeling of fission gas release (FGR) in nuclear fuel performance codes have been relying on empirical models which are not applicable beyond their range of calibration. Recently a more efficient and flexible physics-based FGR and swelling model that can describe a wider range of reactor operation conditions have been implemented in fuel performance code BISON. In this work, we performed Bayesian calibration to inversely quantify the uncertainties associated with the 5 tuning parameters in BISON FGR model, based on the time-dependent FGR measurement data from Riso-AN3 benchmark. It takes around 1 hour and 40 minutes for BISON to simulate the Riso-AN3 benchmark with a moderate mesh size using 32 processors. As hundreds of thousands of BISON simulations are needed during Markov Chain Monte Carlo (MCMC) sampling, running BISON directly is not a viable solution. Therefore we propose to use Kriging-based metamodels to replace BISON during MCMC sampling. Surrogate model is an approximation of the input/output relation of a computer code/model. It is also called metamodel, response surface, emulator. Surrogate models usually take much less computational time than the original model (like BISON) while maintaining the input/output relation of the original model to a desirable accuracy. Constructing the metamodels normally requires small number of executions of the original model. Once validated, metamodels can be used to perform uncertainty and sensitivity analysis, Bayesian calibration, optimization, etc.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Standalone BISON Fuel Performance Results for Watts Bar Unit 1, Cycles 1-3
The Consortium for Advanced Simulation of Light Water Reactors (CASL) is moving forward with more complex multiphysics simulations and increased focus on incorporating fuel performance analysis methods. The coupled neutronics/thermal-hydraulics capabilities within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) have become relatively stable, and major advances have been made in analysis efforts, including the simulation of twelve cycles of Watts Bar Nuclear Unit 1 (WBN1) operation. While this is a major achievement, the VERA-CS approaches for treating fuel pin heat transfer have well-known limitations that could be eliminated through better integration with the BISON fuel performance code. Several approaches are being implemented to consider fuel performance, including a more direct multiway coupling with Tiamat, as well as a more loosely coupled one-way approach with standalone BISON cases. Fuel performance typically undergoes an independent analysis using a standalone fuel performance code with manually specified input defined from an independent core simulator solution or set of assumptions. This report summarizes the improvements made since the initial milestone to execute BISON from VERA-CS output. Many of these improvements were prompted through tighter collaboration with the BISON development team at Idaho National Laboratory (INL). A brief description of WBN1 and some of the VERA-CS data used to simulate it are presented. Data from a small mesh sensitivity study are shown, which helps justify the mesh parameters used in this work. The multi-cycle results are presented, followed by the results for the first three cycles of WBN1 operation, particularly the parameters of interest to pellet-clad interaction (PCI) screening (fuel-clad gap closure, maximum centerline fuel temperature, maximum/minimum clad hoop stress, and cumulative damage index). Once the mechanics of this capability are functioning, future work will target cycles with known or suspected PCI failures to determine how well they can be estimated.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Executing BISON-CASL Fuel Performance Cases Using VERA-CS Output
As the Consortium for Advanced Simulation of Light Water Reactors (CASL) moves forward with more complex multiphysics simulations, there is increased focus on incorporating fuel performance analysis methods. The coupled neutronics/thermal-hydraulics capabilities within the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) have become relatively stable and major advances have been made in analysis efforts, including the simulations of twelve cycles of Watts Bar Nuclear Unit 1 (WBN1) operation. However, VERA-CS approaches for treating fuel pin heat transfer have well-known limitations that could be eliminated through better integration with the BISON-CASL fuel performance code. Several approaches are being taken to improve or replace the VERA-CS fuel models and improve integration with BISON-CASL. Tiamat is being developed to replace the fuel models with a direct coupling of BISON-CASL and the neutronics and coolant heat transfer of VERA-CS. Tiamat will execute in an inline mode that allows for in-memory transfer of VERA-CS data to BISON-CASL. This will provide fuel performance results during core simulation. However, fuel performance typically undergoes an independent analysis using a stand-alone fuel performance code with manually specified input defined from an independent core simulator solution or set of assumptions. This milestone covers initial efforts to facilitate using VERA for core simulation and fuel performance to operate in this mode of execution. This approach will be used for several CASL fuel performance challenges, such as RIAs and DNB. These issues require time-dependent full core results to establish the boundary conditions for the high-fidelity BISON-CASL simulations. The approach will also be used to improve fuel temperature models in VERA-CS, guide the development of Tiamat, and simplify the use of BISON-CASL within VERA for industry users.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
CAMPUS: Fully Coupled Multiphysics Modeling and Simulation of LWR and FBR Nuclear Fuel Behavior
Nuclear fuels operate in an environment that involves complex multiphysics phenomena, and the multiphysics behavior is often tightly coupled. The prediction of fuel behavior under irradiation involves very complex material behavior that is not seen in any other engineering discipline. Understanding and predicting the evolving material properties and phenomenal behaviors of nuclear fuels are major challenges for fuel design and behavior prediction. The multiphysics behavior is often a tightly coupled thermal, mechanical and chemical phenomenon associated with a large number of control parameters and uncertainties, which requires an advanced fuel performance modeling and simulation capability to improve the existing reactors performance, to address environmental and reactor safety concerns. Nuclear fuel performance modeling and simulation capability is necessary as a result of the high cost and difficulty in experimental measurements. So, a fuel performance code called CityU Advanced Multiphysics Nuclear Fuels Performance with User-defined Simulations (CAMPUS) was developed to predict both light water reactor and fast reactor fuel behavior. An effective approach was developed using self-defined multiphysics models based on the framework of COMSOL Multiphysics to manage the nonlinearities associated with light water reactor and fast reactor fuels performance analysis. The modelling and simulations approach is of particular importance since it demonstrates highly interrelated and nonlinear fuel performance models can be developed and implemented into a multiphysics platform, and thus more efforts can be put into physics model development and improvement. The modelling results can be used to identify the important phenomena and parameters, to direct and design experiments and to assist in post analysis of expensive nuclear experiments for light water reactor and fast reactor nuclear fuel performance analysis. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Pellet Cladding Mechanical Interaction Modeling Using the Extended Finite Element Method
As a brittle material, the ceramic UO2 used as light water reactor fuel experiences significant fracturing throughout its life, beginning with the first rise to power of fresh fuel. This has multiple effects on the thermal and mechanical response of the fuel/cladding system. One such effect that is particularly important is that when there is mechanical contact between the fuel and cladding, cracks that extending from the outer surface of the fuel into the volume of the fuel cause elevated stresses in the adjacent cladding, which can potentially lead to cladding failure. Modeling the thermal and mechanical response of the cladding in the vicinity of these surface-breaking cracks in the fuel can provide important insights into this behavior to help avoid operating conditions that could lead to cladding failure. Such modeling has traditionally been done in the context of finite-element-based fuel performance analysis by modifying the fuel mesh to introduce discrete cracks. While this approach is effective in capturing the important behavior at the fuel/cladding interface, there are multiple drawbacks to explicitly incorporating the cracks in the finite element mesh. Because the cracks are incorporated in the original mesh, the mesh must be modified for cracks of specified location and depth, so it is difficult to account for crack propagation and the formation of new cracks at other locations. The extended finite element method (XFEM) has emerged in recent years as a powerful method to represent arbitrary, evolving, discrete discontinuities within the context of the finite element method. Development work is underway by the authors to implement XFEM in the BISON fuel performance code, and this capability has previously been demonstrated in simulations of fracture propagation in ceramic nuclear fuel. These preliminary demonstrations have included only the fuel, and excluded the cladding for simplicity. This paper presents initial results of efforts to apply XFEM to model stress concentrations induced by fuel fractures at the fuel/cladding interface during pellet cladding mechanical interaction (PCMI). This is accomplished by enhancing the thermal and mechanical contact enforcement algorithms employed by BISON to permit their use in conjunction with XFEM. The results from this methodology are demonstrated to be equivalent to those from using meshed discrete cracks. While the results of the two methods are equivalent for the case of a stationary crack, it is demonstrated that XFEM provides the additional flexibility of allowing arbitrary crack initiation and propagation during the analysis, and minimizes model setup effort for cases with stationary cracks.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Pellet-clad mechanical interaction screening using VERA applied to Watts Bar Unit 1, Cycles 1–3
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. In conclusion, 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.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Sensitivity analysis and fuel performance of a control rod withdrawal in the generic fluoride-salt-cooled high temperature reactor
Not Available
Barthle, Jonathan L.↗
Analysis of a Loop for Fission Gas Release and Monitoring at the PULSTAR Reactor
Modern nuclear fuel performance analysis tools aim at establishing a predictive approach for examining macroscopic fuel behavior due to fuel gaseous swelling and the associated pellet cladding mechanical interaction (PCMI). In general, the macroscopic behavior is highly sensitive to understanding microscopically driven phenomena, which in this case is the release of fission gas (i.e., Kr and Xe) from the fuel. Previous studies identified the uncertainties associated with the intra-granular fission gas diffusion coefficient as most limiting for the accuracy of fission gas release calculations. In particular, it was concluded that, in order to improve fission gas release as well as gaseous swelling predictions using fuel performance codes, a better characterization of the diffusion coefficient is needed. This work explores the implementation of an on-line fission gas release and measurement loop at the North Carolina State University (NCSU) PULSTAR reactor. A unique characteristic of this loop is the ability to control the temperature of the fuel samples during irradiation to establish a more accurate correlation between the irradiation conditions and an observable such as the intragranular gas diffusion coefficient. Analysis was performed of an 'in-pile' fission gas release experiment that is currently under implementation at the NCSU PULSTAR reactor. Using MCNP6 calculations and COMSOL simulations the fission rate density and temperature distributions in UO{sub 2} microspheres were estimated. To explore the feasibility of detecting fission gas isotopes, estimates of R/B for {sup 138}Xe were obtained using the equivalent sphere model and combined with MCNP6 predicted B estimates to obtain R. Assuming detection in a typical gamma-ray spectrometry HPGe system, it was conclude that reasonable count rates will be achieved to allow extraction of release information. At this stage, accurate assessment is underway of the performance of various loop components to set the final design criteria in support of loop fabrication. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Uncertainty Quantification and Sensitivity Analysis Applications to Fuel Performance Modeling
Best-estimate fuel performance codes such as BISON currently under development at the Idaho National Laboratory, utilize empirical and mechanistic lower-length-scale informed correlations to predict fuel behavior under normal operating and accident reactor conditions. Traditionally, best-estimate results are presented using the correlations with no quantification of the uncertainty in the output metrics of interest. However, there are associated uncertainties in the input parameters and correlations used to determine the behavior of the fuel and cladding under irradiation. Therefore, it is important to perform uncertainty quantification and include confidence bounds on the output metrics that take into account the uncertainties in the inputs. In addition, sensitivity analyses can be performed to determine which input parameters have the greatest influence on the outputs. In this paper we couple the BISON fuel performance code to the DAKOTA uncertainty analysis software to analyze a representative fuel performance problem. The case studied in this paper is based upon rod 1 from the IFA-432 integral experiment performed at the Halden Reactor in Norway. The rodlet is representative of a BWR fuel rod. The input parameters uncertainties are broken into three separate categories including boundary condition uncertainties (e.g., power, coolant flow rate), manufacturing uncertainties (e.g., pellet diameter, cladding thickness), and model uncertainties (e.g., fuel thermal conductivity, fuel swelling). Utilizing DAKOTA, a variety of statistical analysis techniques are applied to quantify the uncertainty and sensitivity of the output metrics of interest. Specifically, we demonstrate the use of sampling methods, polynomial chaos expansions, surrogate models, and variance-based decomposition. The output metrics investigated in this study are the fuel centerline temperature, cladding surface temperature, fission gas released, and fuel rod diameter. The results highlight the importance of quantifying the uncertainty and sensitivity in fuel performance modeling predictions and the need for additional research into improving the material models that are currently available.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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. Further, 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 UO 2 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.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Fuel Performance Simulation of Iron-Chrome-Aluminum Cladding during Steady-State LWR Operation
During a severe accident scenario, such as a loss-of-coolant accident, a nuclear reactor may lose its capacity to cool its fuel. The fuel rod temperatures begin to increases, quickly reaching temperatures where the coolant begins to evaporate. This lowers the coolant level in the core and eventually uncovers the fuel. Without sufficient heat transfer from the fuel rods to the coolant, the temperature of the fuel and cladding will increase dramatically. In the case of traditional zirconium-based alloys, as the temperature of the cladding reaches ∼1200 deg. C the zirconium rapidly begins to oxidize with H{sub 2}O in the coolant and steam. This oxidation reaction causes both thinning of the Zr-alloy cladding as the metal reacts and releases large amounts of H{sub 2} gas into the reactor pressure vessel. In order to increase the safety margin of LWR fuel in severe accident scenarios, several alternative cladding materials have been proposed to replace the currently used zirconium-based alloys. Of these materials, there is a particular focus on select iron-chrome-aluminum (FeCrAl) alloys because of slower oxidation Kinetics in high temperature steam than zirconium-based alloys. This increased oxidation resistance may provide additional coping time to mitigate any further damage resulting from an accident. However, by changing the cladding material, the neutronic characteristics of the core will change. FeCrAl alloys exhibit thermal neutron absorption cross sections nearly ∼12-16 x greater than Zr-alloys. This decreases the reactivity of the core, subsequently decreasing the potential cycle length for operation. To counteract this, fissile material may be added in the form of increased enrichment or increased fuel mass or the cladding thickness might be decreased, effectively reducing the neutronic penalty from using the alloy. In this study, all three changes are considered. To assess the viability of FeCrAl as an alternative cladding to Zr-alloys, engineering-scale fuel performance analysis is used to monitor the cladding and integral fuel rod behavior through its operation. Ongoing work is being performed to identify key parameters essential to modeling FeCrAl alloys for LWR applications and properly model fuel compliance and stress relief mechanisms for UO{sub 2} fuel. To improve the predictive capability of fuel performance codes for FeCrAl cladding under normal operation, high-fidelity mechanistic/physics-based material models are needed. This information can be generated and improved by access to experimental data on targeted FeCrAl alloys. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Ceramic nuclear fuel fracture modeling with the extended finite element method
Ceramic fuel pellets used in nuclear light water reactors experience significant fracture due to the high thermal gradients experienced under normal operating conditions. This has important effects on the performance of the fuel system. Because of this, a realistic, physically based fracture modeling capability is essential to predict fuel behavior in a wide variety of normal and off-normal conditions. The extended finite element method (X-FEM) is a powerful method to represent arbitrary propagating discrete cracks in finite element models, and has many characteristics that make it attractive for nuclear fuel performance analysis. This paper describes the implementation of X-FEM in a multiphysics fuel performance code and presents applications of that capability. These applications include several thermal mechanics fracture benchmark problems, which demonstrate the accuracy of this approach. It also includes application of this capability to study nuclear fuel fracture, both on stationary and propagating cracks. The study on stationary cracks shows the effects that varying the number of radial cracks and the length of those cracks have on the energy release rate. Here, the propagating crack case demonstrates random initiation and subsequent propagation of interacting thermally induced cracks during an initial ramp to full power with fresh fuel.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗