Accident tolerant fuels (FeCrAl Cladding & Coating) performance analysis in Boiling Water Reactor (BWR) by the MELCOR 1.8.6 UDGC
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Wang, Jun (ORCID:0000000245504603)↗
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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↗
Vehicle Design Evaluation Program (VDEP). A computer program for weight sizing, economic, performance and mission analysis of fuel-conservative aircraft, multibodied aircraft and large cargo aircraft using both JP and alternative fuels
The NASA Langley Research Center vehicle design evaluation program (VDEP-2) was expanded by (1) incorporating into the program a capability to conduct preliminary design studies on subsonic commercial transport type aircraft using both JP and such alternate fuels as hydrogen and methane;(2) incorporating an aircraft detailed mission and performance analysis capability; and (3) developing and incorporating an external loads analysis capability. The resulting computer program (VDEP-3) provides a preliminary design tool that enables the user to perform integrated sizing, structural analysis, and cost studies on subsonic commercial transport aircraft. Both versions of the VDEP-3 Program which are designated preliminary Analysis VDEP-3 and detailed Analysis VDEP utilize the same vehicle sizing subprogram which includes a detailed mission analysis capability, as well as a geometry and weight analysis for multibodied configurations.
Oman, B. H.↗
The MELCOR Analysis of Chinshan Nuclear Power Plant Spent Fuel Pool for Fukushima-like Accident
The safety analysis of the nuclear power plant (NPP) is very important work in the NPP safety. Especially after the Fukushima NPP accident occurred, the importance of NPP safety analysis has been raised and there is more concern for the safety of NPPs in Taiwan. Because the earthquake and tsunami occurred, the cooling system of spent fuel pool failed and the safety issue of spent fuel pool generated in Japan's Fukushima NPP. Chinshan NPP is the first NPP in Taiwan which is BWR/4 plant and OLTP (Original Licensed Thermal Power) for each unit is 1775 MWt. Chinshan NPP finished SPU (stretch power uprate) and the operating power is 103.66% of the OLTP, which is 1840 MWt now. After Fukushima NPP accident occurred, in order to concern the safety of Chinshan NPP spent fuel pool, we performed the safety analysis of spent fuel pool by using MELCOR/SNAP which also assumed the cooling system of spent fuel pool failed. The geometry of Chinshan NPP spent fuel pool is 12.17 m x 7.87 m x 11.61 m and the initial condition is 60 deg. (water temperature) / 1.013 x 10{sup 5} Pa. And, the total power of the fuels is roughly 8.9 MWt initially. MELCOR is developed by Sandia National Laboratories. MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor NPPs, including containment response and leakage of activity to the environment. A broad spectrum of severe accident phenomena in both BWRs and PWRs is treated in MELCOR in a unified framework. These include thermalhydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. According to Dr. Carbajo's paper, MELCOR was used to perform the study of spent fuel pool of Fukushima Daiichi Unit 4. Niina E. Koenoenen used MELCOR to establish the spent fuel pool model of a Nordic BWR. The loss-of-pool-cooling accidents was simulated and analyzed by this model. The above studies indicate that MELCOR is capable of handling the simulation of the spent fuel pool. An increasing number of researchers are using MELCOR code to analyze test facilities and nuclear power plants. Jun Wang et al. used MELCOR to perform core degradation simulation and MELCOR results were compared with CORA experimental data. Through this work, they reviewed the performance of MELCOR COR package in detail. Tae Woon Kim et al. developed MELCOR model of APR1400 (an advanced pressurized water reactor). They used this model to perform the LBLOCA transient simulation and sensitivity study. Longze Li et al. established the MELCOR model of CPR1000 (a Chinese pressurized reactor 1000-MW power plant). The station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV) transient was simulated and analyzed by this model. SNAP is a graphic user interface program that processes the inputs and outputs of MELCOR. In addition, there is an animation function in SNAP which can present the animation of analysis results. There are two main steps in this research. The first step was the establishment of Chinshan NPP spent fuel pool MELCOR/SNAP model. And the transient analysis of Chinshan NPP spent fuel pool MELCOR/SNAP model under the cooling system failure condition was performed. The next step was the comparison among the MELCOR, TRACE and CFD data. TRACE and CFD data were from INER reports. In addition, the animation model of Chinshan NPP spent fuel pool was presented by using the animation function of SNAP with MELCOR/SNAP analysis results. This study has developed MELCOR/SNAP model of Chinshan NPP spent fuel pool successfully. By using the above model, the safety analysis of the spent fuel pool was performed under the cooling system of spent fuel pool failed condition. The analysis results of MELCOR, TRACE and CFD were similar in this case. It indicated that there was a respectable accuracy in MELCOR/SNAP model. The analysis results depicted that the uncovered of the fuels occurred at 2.6 day and the metal-water reaction of fuels occurred roughly at 3.7 day after the cooling system failed. The above results indicated that the failure of cladding occurred after 3.7 day. This study's results can help to evaluate the safety issue of Chinshan NPP spent fuel pool. In addition, the cooling system and some mitigation measures of spent fuel pool is not simulated in the above model. However, we will do it in the future. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Uncertainty and Sensitivity Analysis of Reactivity Insertion Accidents using BISON Fuel Performance Code
The development of Accident Tolerant Fuels (ATF) has expanded into a considerable research program in the wake of the events at the Japanese Fukushima Daiichi nuclear power plant in 2011. The ATF goal is to insert a test assembly into a commercial reactor by 2022. To support this, testing of the ATF design under off-normal conditions is needed to assess the 'enhanced accident tolerance' capability of the fuel rod design over standard UO{sub 2}-Zircaloy concepts. Transient testing capabilities are necessary for off-normal testing of events such as loss of coolant accidents (LOCA) and reactivity insertion accidents (RIA). A transient testing program has been developed to support the testing of ATF designs. The Transient Reactor Test (TREAT) facility has been chosen as the site to perform the transient testing of the ATF designs. TREAT is a versatile test facility capable of subjecting fuel specimens to various off-normal conditions. Historic TREAT experiments were performed in a variety of experimental test vehicles including static environment capsules, once-through flowing systems, and also recirculating flowing systems. The planned TREAT experiments will be operated under PWR conditions. Therefore, a static capsule vehicle is under development to handle required temperatures and pressures. The static capsule, termed the Static Environment Rodlet Transient Test Apparatus (SERTTA), will be used for irradiation tests in TREAT to subject ATF fuel rodlets to a range of power pulse conditions and environments. To aid in the design and analysis of future tests in TREAT, the fuel performance code BISON is being used. Fuel performance codes capable of accident modeling have been developed for some time and they have shown an ability to reproduce experimental results with some degree of adequacy, but they still require extensive verification and validation. Future experiments in TREAT will add to a limited database of past transient tests performed in transient reactors to support LWR systems. These tests can be used for future verification and validation of fuel performance codes. Understanding the impact of the uncertainty of influential parameters is critical for not only analyzing the results of the experiment but also for modeling and validation of codes. For example, uncertainties associated with the initial manufactured state of the sample, instrument measurements, material property models, and planned experiment conditions can provide important insight into model performance, experimental data, and play an important role in experimental design. The objective of this work is to perform an uncertainty and sensitivity analysis on a simplified RIA case of a standard UO{sub 2}-Zircaloy sample using expected nominal conditions in the SERTTA capsule in TREAT. The BISON fuel performance code is used in conjunction with DAKOTA (Design Analysis Kit for Optimization and Tera-scale Applications) to perform the uncertainty and sensitivity analysis. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Demonstration of the Coating Analysis Capabilities in BISON
In order to improve the accident tolerance of light water reactor (LWR) fuel, coated cladding materials have been proposed which exhibit much slower oxidation kinetics in high-temperature steam than Zr-alloys. This behavior 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. In order to determine how these coating materials will adjust the behavior of the Zr-alloy cladding during normal reactor operation, simulations have been performed using the Bison fuel performance code. This analysis focused on modeling and comparing the integral thermo-mechanical performance of coated and uncoated zircaloy cladding with uranium dioxide fuel under normal operating conditions. This analysis demonstrates the ability to model a discrete meshed coating bonded to the cladding exterior and provides a comparison with an uncoated fuel rod. A generic range of coating material properties are implemented given the differences and unknowns regarding the proprietary coating properties used by the vendors. A main-effects analysis was performed to identify trends and sensitivities in how coating material properties affect key fuel performance metrics. Lastly, because residual stresses are expected to form in the coating, arising from the high temperature coating application process and application techniques, an additional analysis was performed to identify the impact of these stresses on the cladding behavior. Future analysis efforts will target more comprehensive documentation of this main effects analysis by sampling the results at the beginning and middle-of-life in order to capture important behavior before and during the onset of gap closure.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
PCI margin analysis methodology using the FALCON fuel performance code
ANATECH Corp., a wholly owned subsidiary of Structural Integrity Associates, Inc. has been at the forefront of PCI (pellet-cladding interaction) margin analyses as principal author of the EPRI PCI Guidelines documents and through the performance of numerous PCI margin evaluations for light-water reactors. In this paper, ANATECH first provides a discussion of specific features of the methodology developed over a number of years of application including a) selection of limiting assemblies for the detailed FALCON full length fuel rod (i.e., R-Z model) and PCI (i.e., R-θ model) evaluations, b) development of FALCON input characteristics including high fidelity rod-by-rod power histories, and c) application of failure criteria and remedies to the core designer in the case where the criteria are exceeded. Second, ANATECH provides guidance based on evaluations performed to-date. Analysis results are presented to show critical aspects of the evaluations and how they are used to provide insights to inform and improve commercial plant operation.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Monte Carlo fuel burnup analyses of the Accelerator-Driven Subcritical systems
Here, we discuss the Monte Carlo fuel burnup analysis performed for the Argonne National Laboratory (ANL) Accelerator-Driven Subcritical (ADS) conceptual design. The ANL ADS conceptual design was determined to dispose of the 131 tons of Minor Actinides (MA) fuel of the current United States Spent Nuclear Fuel (SNF) inventory. It uses a neutron source from 1 GeV 25MW proton beam with the liquid lead target to drive the subcritical fission assembly. It utilizes the liquid metal slurry fuel concept to load the MAs fuel. The subcritical fission assembly has a neutron multiplication factor of 0.98. The total fission power generated in the system is about 2.67 GW. The system is designed to operate for 35 full power years. Monte Carlo fuel burnup analyses were performed to calculate the amount of MAs consumed by the ANL ADS conceptual design. The MCNP6 computer program was used to simulate the high energy physics of the protons striking the liquid lead target for generating the neutron source. The MCB5 and SERPENT Monte Carlo burnup computer programs were used to simulate the fuel burnup during the ADS operation. In the ANL ADS conceptual configuration, the reactivity loss during the fuel burnup is compensated by adding fresh fuel into the subcritical assembly. In this paper, both batch and continuous fueling modes are examined. The fresh fuel particles are loaded into the slurry fuel carrier at the beginning of each burnup step of the batch-fueling mode and loaded continuously into the slurry fuel carrier during the operation in the continuous fueling mode. Monte Carlo fuel burnup analyses show that one ANL ADS conceptual design can consume about 25.7 tons of MAs fuel during the 35 full power years with the continuous fueling mode. About five units is required to dispose of the 131 tons of MAs
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Dynamical System Scaling Application to Zircaloy Cladding Thermal Response During Reactivity-Initiated Accident Experiment
New fuel design and development currently requires 20 to 25 years to be qualified for use by the nuclear power industry. The thermal-hydraulics community has taken advantage of scaling theory to design reduced scale experiments that correctly preserve dominant key phenomena while quantifying distorted phenomena. These techniques can be leveraged in the design and analysis of fuel performance experiments to help reduce the timeline associated with fuel design and development. This study uses the Dynamical System Scaling (DSS) method to analyze cladding temperature data from the recent SETH-C experiment in the TREAT facility and accompanying BISON simulations to assess dynamic distortions occurring throughout the fast power excursion transient. The DSS analysis revealed that on the cool down from peak cladding temperature that the fuel radial power profile is the most sensitive modeling parameter with a heterogenous radial peaking factor corresponding to the lowest distortion compared to a uniform energy deposition. For the heat up to peak cladding temperature the heterogeneous radial power profile corresponded to the shortest process action. Finally, for the heat up to peak cladding temperature, the gap conductance model sensitivity was quantified using process action and shows that the default Light Water Reactor gap conductance model corresponded to the longest process action.
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↗
Validation of MFUEL Metal Fuel Performance Models of SAS4A/SASSYS-1
The fuel characterization models of SAS4A/SASSYS-1 (SAS) have recently been extended to include a new U-Pu-Zr metal fuel model, MFUEL. MFUEL is equipped with mechanistic physics based models to predict the pre-transient characterization and transient response of metal fuel, with emphasis on fuel melting, cladding failure, and the metal fuel’s impact on core reactivity. The MFUEL model will be available in the full version of SAS4A/SASSYS-1 5.7, which is scheduled to be released in June 2023. Fast reactor fuel pins that operated in EBR-II and FFTF with low smear density U-Zr and U-Pu-Zr metal fuels and irradiation resistant ferritic-martensitic cladding showed significant advantages in achieving high burnups and assuring inherent safety characteristics during anticipated transients, design basis events, and beyond design basis events. For safety analysis, a fuel performance model must be able to predict (1) Fuel pin mechanics and compositional and dimensional changes, (2) Clad failure, and (3) Fuel pin thermal resistance. Achieving these high level goals accurately is strongly related to the model performance of individual physical processes taking place within a fuel pin during its lifetime. Metal fuels typically operate above the mid-point temperature of melting during normal, as well as off-normal, conditions. At these elevated temperatures, the availability of thermal activation provides a driving force for various diffusional processes leading to complex phase transformations, micro-structure evolution, significant amounts of fuel swelling, interconnected porosity formation, excessive amounts of fission gas release, and fuel clad chemical interactions. Clad failure in fast reactors primarily occurs as a result of creep rupture augmented by clad wastage formation. The reaction is driven by thermal creep induced dislocation motion, grain boundary cavity nucleation, growth and breakup of grain boundaries. The high level complexity and limited available data requires introducing physics-based modeling approaches to gain extrapolation ability and sensitivity with respect to various conditions. The objective of this report is to perform validation of MFUEL using the experimental data for (1) Normal operation EBR-II fuel behavior, (2) Normal operation PHENIX fuel behavior, (3) HT9 Pressure tube ramp-and-hold creep rupture tests, (4) Whole Pin Furnace (WPF) creep strain, creep rupture and eutectic tests, (5) Fuel Behavior Test Apparatus (FBTA) eutectic tests, and (6) TREAT M5-7 OverPower tests up to clad failure. Section-2 includes a brief description of the MFUEL models. A detailed description of the MFUEL physics-based, semi-empirical models will be presented in the SAS V 5.7 theory manual. Section-3, 4, and 5 describes the validation effort for the pre-transient irradiation, furnace transients, and TREAT M-Series transients, respectively.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Modeling, Performance Assessment, and Nodal Data Analysis of TRISO-Fueled Systems with Shift
This technical report documents several enhancements to the Shift Monte Carlo (MC) code under the US Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in fiscal year (FY) 2022. Performance enhancements were added to Shift specifically for tristructural isotropic (TRISO)–fueled reactor systems and guided based on performance analysis in FY 2021. For the pebble performance model developed in previous studies, the runtime improved by ~ 91× compared to the original model and ~ 2× compared to the user-optimized model. Compared to Serpent, Shift is ~ 3× slower if Serpent delta-tracking is enabled but ~ 2× faster when delta-tracking is disabled. The multigroup cross section generation was improved through simplifying tally input definitions, porting several post-processing tally operations from Python scripts into the Shift code base, and accounting for production reactions in the scattering multiplicity. Progress was also made on two emerging capabilities: (1) the development of Titan (a Shift reactor physics user interface) and (2) initial investigation into path-length tallies for computing multigroup scattering matrices.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Pressurized-Water Reactor Core Design Demonstration with Genetic Algorithm Based Multi-Objective Plant Fuel Reload Optimization Platform
LWRS M3 milestone report due September 15, 2023. This report summarizes development and demonstration activities of PRLO optimization platform built in Risk Analysis and Virtual ENviroment (RAVEN), specifically: (1) Improvement of multi-objective non-dominated sorting genetic algorithm II (NSGA-II) to handle large size of objectives and constraints, (2) Demonstration of pressurized water reactor core design with NSGA-II multi-objective optimization platform, and (3) Single-objective optimized core design including system analysis and fuel performance feedback.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
MV Sea Change: Fuel Cell, Emissions, and Hydrogen Fueling Performance
The Sea Change Project, managed and financed by SWITCH Maritime, produced the first zero-emission 100% hydrogen fuel-cell powered commercial ferry in the world. The MV Sea Change is a 70-foot catamaran equipped with a 360-kW fuel cell powertrain and a 250-bar 246 kg capacity hydrogen storage system. The 75-passenger vessel began public passenger service in July 2024 as part of the San Francisco Bay Ferry system. Summarized here are aspects of the design of the vessel, its construction, and vessel bunkering with high-pressure hydrogen. Results from a fuel cell performance analysis are also presented, along with estimates of the greenhouse gas (GHG) and criterial pollutant emissions assuming different sourcings of hydrogen. Lessons learned from the project will be briefly reviewed.
08 HYDROGEN↗
Bison Methods and Contact Algorithms
A review of methods in Bison, including contact capabilities.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Application of sensitivity analysis in DYMOND/Dakota to fuel cycle transition scenarios
The ability to perform sensitivity analysis has been enabled for the nuclear fuel cycle simulator DYMOND through its coupling with the design and analysis toolkit Dakota. To test and demonstrate these new capabilities, a transition scenario and multi-parameter study were devised. The transition scenario represents a partial transition from the US nuclear fleet to a closed fuel cycle with small modular LWRs and fast reactors fueled by reprocessed used nuclear fuel. Four uncertain parameters in this transition were studied – start date of reprocessing, total reprocessing capacity, the nuclear energy demand growth, and the rate at which the fast reactors are deployed – with respect to their impact on four response metrics. The responses – total natural uranium consumed, maximum annual enrichment capacity required, total disposed mass, and total cost of the nuclear fuel cycle – were chosen based on measures known to be of interest in transition scenarios and to be significantly impacted by the varying parameters. Furthermore, analysis of this study was performed both from the direct sampling and through surrogate models developed in Dakota to calculate the global sensitivity measures Sobol’ indices. This example application of this new capability showed that the most consequential parameter to most metrics was the share of new build capacity that is fast reactors. However, for the cost metric, the scaling factor of the energy demand growth was significant and had synergistic behavior with the fast reactor new build share.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
The impact of interface bonding efficiency on high-burnup spent nuclear fuel dynamic performance
Finite element analysis (FEA) was used to investigate the impact of interfacial bonding efficiency at pellet-pellet and pellet-clad interfaces of high-burnup (HBU) spent nuclear fuel (SNF) on system dynamic performance. Bending moments M were applied to FEA model to evaluate the system responses. From bending curvature, κ, flexural rigidity EI can be estimated as EI = M/κ. The FEA simulation results were benchmarked with experimental results from cyclic integrated reversal bending fatigue test (CIRFT) of HBR fuel rods. The consequence of interface debonding between fuel pellets and cladding is a redistribution of the loads carried by the fuel pellets to the clad, which results in a reduction in composite rod system flexural rigidity. Furthermore, the interface bonding efficiency at the pellet-pellet and pellet-clad interfaces can significantly dictate the SNF system dynamic performance. With the consideration of interface bonding efficiency, the HBU SNF fuel property was estimated with CIRFT test data.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗