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This report summarizes the FY2022-23 work performed by Argonne National Laboratory (ANL) on sibling pins that are part of the High Burnup (HBU) Spent Fuel Data Project. The project is funded by the US Department of Energy’s Spent Fuel & Waste Science and Technology (SFWST) Office, which is part of the Office of Nuclear Energy’s Spent Fuel and Waste Disposition Office.
The United States (U.S.) nuclear industry faces a challenge in maintaining required levels of safety while ensuring economic competitiveness to stay in business. Safety remains a key parameter for all aspects of light water reactor (LWR) nuclear power plant (NPP) operations. Safety can become more economical by using a risk-informed ecosystem, such as the one being developed by the Risk-Informed Systems Analysis (RISA) Pathway under the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program. The LWRS Program promotes a wide range of research and development activities with the goal of maximizing both the safety and economic efficiency of NPPs through improved scientific understanding, especially given many plants are now considering second license renewals. The RISA Pathway has two main goals: (1) deploy methodologies and technologies that better represent safety margins and cost and safety factors and (2) develop advanced applications that enable cost-effective plant operation. The Plant Reload Optimization Platform development project aims to build a reactor core design tool that includes reactor safety and fuel performance analyses, and also uses artificial intelligence to support optimization of core design solutions. This report summarizes Fiscal Year 2022 (FY-22) activity in platform capability developments in RAVEN. This platform performs simulations using industry codes for core design (i.e., PARCS) and fuel performance (i.e., TRANSURANUS) which will allow expansion of the capabilities to include advanced fuel designs such as accident-tolerant fuel (ATF)s with high burnup.
The transient fission gas release (tFGR) during the temperature ramp associated with a loss-of-coolant accident (LOCA) in light-water reactors (LWRs) is likely a significant contribution to the total pressure in a fuel rod and may cause an unexpected rod burst. The lack of data related to tFGR continues to be a key gap in understanding LWR cladding burst behavior under LOCA conditions. To fully characterize this behavior, tFGR data must be collected from several different systems that can capture all relevant testing conditions. Oak Ridge National Laboratory (ORNL) has developed a system to measure the integral tFGR from irradiated fuel segments. This system was designed to integrate with the existing Severe Accident Test Station (SATS) and to build upon decades of experience capturing fission gas to characterize fuel behavior. The tFGR system consists of a sweep gas system to transport gases from the in-cell SATS apparatus to an out-of-cell fission gas detection system composed of a series of cold traps to capture the off-gas from the heating tests and a gamma spectrometry system to detect and measure 85 Kr. Initial system testing operations were completed during which 85 Kr collection and measurement were verified along with the ability to detect stable inert gases. A tFGR test with a high-burnup fuel specimen was successfully conducted by the in-cell SATS-tFGR system at ORNL. The posttest examination is under way, the result of which will be reported in FY24.
This report summarizes efforts within NEAMS to investigate the mechanisms that govern fission gas behavior in UO 2 . In particular, the focus is on understanding how fission gas behavior causes transient fission gas release and fragmentation/pulverization of high burnup structure (HBS) in UO 2 . HBS forms in the periphery of the pellet where temperatures are relatively low. Previously, MD simulations were performed to determine the reaction energies for various Xe and U defects with bubbles, as a function of Xe to vacancy ratio or, equivalently, pressure. As had been shown in FY22, it was found that the unmodified version of the Simple Integrated Fission Gas Release and Swelling (SIFGRS) model within BISON greatly over-predicted the number gas atoms per vacancy in the bubbles in the outer rim of the pellet (a ratio of > 1 million). This was due to slow grain boundary vacancy diffusivity and not accounting for the pressure-dependent reaction energy for Xe interstitials with bubbles. The application of the pressure dependent reaction energies was able to restrict Xe to vacancy ratios to 2:1, which is far more realistic than those originally obtained from SIFGRS.
Spent nuclear fuel (SNF) assemblies must be shipped to other sites for processing and disposal. During shipment, the fuel is typically oriented horizontally, and the fuel rods are subject to periodic alternating loads related to the movement of the vehicle that results in alternating bending of the SNF fuel rods. The number of bending cycles is related to the length of the shipping route, with longer routes producing more cycles. Because it is well-known that cyclic loads can produce failures, even when the stress and strain imposed are below the material’s yield point, investigation of the SNF’s fatigue behavior is prudent. This report discusses the results of fatigue testing conducted at Oak Ridge National Laboratory (ORNL) using the Cyclic Integrated Reversible-Bending Fatigue Tester (CIRFT) created by the US Department of Energy Office of Nuclear Energy for the High Burnup Spent Fuel Data Project and its sister rods. The results of the tests are compared with results obtained over the last decade using the same CIRFT for the US Nuclear Regulatory Commission (NRC).
The purpose of this report is to document updates on the apparatus to simulate commercial vacuum drying procedures at the Nuclear Energy Work Complex at Sandia National Laboratories. Validation of the extent of water removal in a dry spent nuclear fuel storage system based on drying procedures used at nuclear power plants is needed to close existing technical gaps. Operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system during subsequent storage and disposal. A general lack of data suitable for model validation of commercial nuclear canister drying processes necessitates well-designed investigations of drying process efficacy and water retention. Scaled tests that incorporate relevant physics and well-controlled boundary conditions are essential to provide insight and guidance to the simulation of prototypic systems undergoing drying processes. This report documents fiscal year 2023 (FY23) updates on the Advanced Drying Cycle Simulator (ADCS). This apparatus was built to simulate commercial drying procedures and quantify the amount of residual water remaining in a pressurized water reactor (PWR) fuel assembly after drying. The ADCS was constructed with a prototypic 17×17 PWR fuel skeleton and waterproof heater rods to simulate decay heat. These waterproof heaters are the next generation design to heater rods developed and tested at Sandia National Laboratories in FY20. In FY23, a series of four simulated commercial drying tests was completed. This report presents the temperature and pressure histories of the drying tests as well as axial temperature profiles that can be compared to data from the Electric Power Research Institute (EPRI) High Burnup Demonstration TN-32B cask. Water content measurements and dew point calculations from a Hiden Analytical HPR-30 mass spectrometer are also presented in this report. Due to familiarization with this first-of-a-kind system, refinements to equipment calibration and test procedures have been identified to better match commercial drying cycles for future simulated tests. However, the presented data demonstrate the successful construction and operation of a viable research platform for quantifying residual water content closely approaching that expected in dry storage canisters during commercial drying procedures.
The Nuclear Energy Agency (NEA) Framework for Irradiation Experiments (FIDES) program includes the High burnup Experiments for Reactivity initiated Accident (HERA) Joint Experimental Program (JEEP). The HERA project is focused on studying Light Water Reactor (LWR) fuel behavior during Reactivity Initiated Accident (RIA) conditions. The HERA experiment plan includes analytical integral experiments using test specimens tailored to investigate specific conditions of relevance as well as prototypic integral experiments focused on irradiated fuel from prototypic origin. Modeling & simulation (M&S) is a key component of any experiment program, and the HERA JEEP is coordinating a M&S exercise. The purpose of this document is to provide problem descriptions to support the HERA M&S exercise based on fuel performance modeling. The HERA M&S exercise is expected to evolve into multiple efforts in outyears. This document may be revised and expanded to incorporate those evolutions.
The United States Department of Energy and the European Atomic Energy Community (EURATOM) collaborate to test, evaluate, and improve the performance of the FRAM 6.1 code with various nuclear materials, especially mixed oxide (MOX) containing an isotopic blend of uranium and very high burnup plutonium. This joint effort is intended to promote effective implementation of the FRAM 6.1 code into nuclear material verification methods.
During a loss-of-coolant accident, a nuclear fuel rods undergo balloon and burst. High burnup fuel is known to fragment and pulverize leading to the potential for the fuel to disperse through the burst opening either during the initial burst or following the burst during the reflood event. This milestone summarizes the design and assembly of a system to study dispersal in a postburst event. Surrogate fuel material (HfO 2 , tungsten, and yttria-stabilized zirconia [YSZ]) dispersing from as-burst loss-of-coolant accident (LOCA) test specimens was extensively investigated, along with dispersal of actual nuclear fuel following a LOCA Severe Accident Test Station (SATS) test.
This report documents the installation and demonstration of computer numerical control (CNC) machining capabilities in the Irradiated Fuels Examination Laboratory hot cell facility at Oak Ridge National Laboratory (ORNL). A modified Tormach PCNC 440 mill was integrated into the hot cell with custom fixturing, fines management, and manipulator-compatible interfaces to enable the fabrication of axial tension test (ATT) and ring tension test (RTT) specimens from irradiated cladding. The first irradiated specimens machined included ATT and RTT geometries harvested from the high-burnup 6XV fuel rod. Dimensional inspections confirmed that machined specimens met the ±0.025 mm tolerance envelope established in prior development; deviations were consistent with expected measurement scatter and inherent specimen variability, such as wall thickness gradients and eccentricity. Comparisons with out-of-cell metrology confirmed that in-cell machining performance aligns with baseline scatter observed under ideal conditions. This work establishes reproducible, end-to-end specimen preparation at ORNL, directly supporting the US Department of Energy’s Accident-Tolerant Fuel program by enabling reliable, traceable mechanical testing of irradiated cladding.
During a loss-of-coolant accident, a nuclear fuel rods undergo balloon and burst. High burnup fuel is known to fragment and pulverize leading to the potential for the fuel to disperse through the burst opening either during the initial burst or following the burst during the reflood event. This milestone summarizes the design and assembly of a system to study dispersal in a postburst event. Surrogate fuel material (HfO2, tungsten, and yttria-stabilized zirconia [YSZ]) dispersing from as-burst loss-of-coolant accident (LOCA) test specimens was extensively investigated, along with dispersal of actual nuclear fuel following a LOCA Severe Accident Test Station (SATS) test. Surrogate material tests spanned a variety of burst geometries, particle types, particle size distributions, and rod oscillation loadings (frequency and amplitude combinations).
This report documents the interim experimental status of small-scale drying tests currently being performed at Pacific Northwest National Laboratory. These small-scale drying tests are to develop the technical bases for sensors, techniques, and approaches that will be used to address the challenges faced when determining moisture content from gas samples taken at the North Anna Nuclear Generating Station for the High Burnup Demonstration Project. Specifically, the purpose of these tests is to support Sandia National Laboratories gas sample methods and analyses through a series of small-scale experiments. Key focus areas include performing tests to: • Correlate sample bottle measurements to in-cask conditions • Estimate the amount of trapped and absorbed water on cask features and surfaces that can contribute to water vapor • Assess the accuracy of the hygrometer used in the measurements under similar conditions • Identify additional methods for measuring humidity in casks easily and accurately using lessons-learned from Sandia National Laboratories, including direct gravimetric and isotopic tracer techniques. To address these topics this report outlines a series of five tests and any data collected to date that were planned for FY 2020 that include 1) humidity measurement method development, 2) gas bottle sampling tests, 3) surface drying tests on cladding, 4) small-scale drying of cladding, and 5) small-scale drying of guide-tube/dashpots. This interim report documents the experimental background, design and progress of these small-scale drying tests. Further data gathered will be summarized in the FY 2021 report.
The extensively evaluated and consistent thermodynamic database, the Molten Salt Thermal Properties Database—Thermochemical (MSTDB-TC), was used along with additional thermodynamic values from other sources as examples of ways to examine molten salt reactor (MSR) fuel behavior. Relative stability with respect to halide potential and temperature for likely fuel and fission product components were mapped in Ellingham diagrams for the chloride and fluoride systems. The Ellingham diagrams provide a rich, visual means for identifying halide-forming components in proposed fuel/solvent salt systems. Thermochemical models and values from MSTDB-TC and ancillary sources were used in global equilibrium calculations to provide compositions for a close analysis of the behavior of a possible Molten Chloride Salt Fast Reactor and a Molten Salt Reactor Experiment-type system at high burnup (100 GWd/t). The results illustrated the oxidative nature of burnup in MSRs and provided information about redox behavior and possible control.
TRistructural ISOtropic (TRISO) particle fuel is a type of nuclear fuel known for its high-temperature and high-burnup performance. Each sub-millimeter diameter TRISO particle consists of uranium-oxycarbide (UCO) or UO2 fuel kernel, coated with buffer, inner pyrolytic carbon (IPyC), silicon carbide (SiC), and outer pyrolytic carbon (OPyC) layers. The SiC layer acts as the main containment barrier for the TRISO particle to retain the fission products, while the IPyC and OPyC layers provide additional barriers to the release of fission products, especially fission gases. During irradiation, phenomena like kernel swelling, buffer densification, and IPyC fracture may impact fuel performance. Post-irradiation microscopy on entire compact cross sections or samples of individual particles deconsolidated from compacts is often used to identify these irradiation-induced changes in morphology. However, each fuel compact generally contains thousands of TRISO particles. To get statistical information on these phenomena, it is cumbersome work if done manually. For example, to get information about swelling/densification behaviors of different layers or kernels after irradiation, researchers previously manually measured the perimeter of each TRISO layer in hundreds of particles after four rounds of iterative grinding and polishing encompassing more than 2000 cross-section images for a total of four fuel compacts. To attempt to reduce the subjectivity inherent in that process and accelerate data analysis, we conducted a study on the automatic TRISO layer segmentation on cross-sectional microscopic images using Convolutional Neural Networks (CNNs). CNNs are a class of machine learning algorithms specifically designed for processing structured grid data that have gained popularity in recent years due to their remarkable performance in various computer vision tasks, including image classification, object detection, and image segmentation. In this research, we have generated the large irradiated TRISO layer dataset with more than 2000 cross-section TRISO microscopic images and the corresponding annotated images. Based on these annotated images, we have employed different CNNs for automatic segmentation of different TRISO layers. These include RU-Net (developed in this study), as well as three existing architectures: U-Net, Residual Network (ResNet), and Attention U-Net. The preliminary results show that the model based on RU-Net has the best performance in terms of intersection-over-union (IoU). Through the aid of these CNN models, we can expedite the analysis of TRISO particle cross-sections, significantly reducing the manual labor involved and improving the objectivity of the segmentation results.
Over the years of commercial nuclear power plant operations in the United States (U.S.), many fuel assemblies have lost the capability to perform all of their desired functions. These assemblies have lost the capability to be handled, stored, or transported to meet regulations resulting in them being classified as damaged fuel. The causes of these failed assemblies are diverse and plant-specific. The majority of these failed assemblies are contained in a damaged fuel can that will be used in conjunction with a storage and/or transportation system. The storage and transportation systems have a limit on how many slots can be filled with damaged fuel cans. A damaged fuel can is generally a stainless steel container that confines damaged SNF and is closed on its end by screened openings. These screened openings allow gaseous and liquid media to escape but minimize the dispersal of gross particulate material. Out of an abundance of caution a few reactors have loaded high burnup fuel into damaged fuel cans. Damaged fuel is not licensed for storage or transport in the U.S., because the regulations for storage and transport do not specify exactly how to classify damaged fuel. Instead the regulations license/certify packages that specify approved contents. Damaged fuel must be included in the approved contents to be a viable option. In many cases damaged SNF is encapsulated in a damaged fuel can to ensure it can confine gross fuel particles, debris, and or damaged assemblies to a known volume within a loaded cask. This damaged fuel can may then be utilized in the same way as an assembly in a storage and transportation design system. Some storage cask systems utilize top and bottom plugs to confine debris in damaged fuel. The most recent domestic documentation on damaged SNF was performed by the U.S. Energy Information Administration (EIA) and used data from U.S. reactors concluding in June 30th 2013 to produce Form GC-859, “Nuclear Fuel Data Survey”. Based on this form there were 136,821 boiling water reactor (BWR) spent nuclear fuel (SNF) assemblies and 104,647 pressurized water reactor (PWR) SNF assemblies for a total of 241,468 SNF assemblies in the U.S. Out of these 241,468 SNF assemblies, 4,521 assemblies were classified as failed. Some assemblies were also disassembled and the fuel rods or pieces of fuel rods were combined together to make a consolidated assembly. A consolidated assembly may include damaged fuel, or it could have been consolidated as part of a demonstration project. The GC-859 data includes 2,550 consolidated assemblies containing between 0 and 264 whole fuel rods. These consolidated assemblies could be placed in single assembly canisters and stored in the spent fuel pool. For dry storage and transportation, a single assembly canister is generally placed in a damaged fuel can. In addition to the consolidated assemblies, 2,391 uncanistered fuel rod pieces exist, which were removed from 494 assemblies.
This report describes the results from electrochemical corrosion tests that were conducted using simulated spent fuel (SIMFUEL) materials comprised of UO2 and surrogate fission products. Two SIMFUEL compositions were tested to quantify the effect of noble metal inclusions and dissolved H2 on the UO2 dissolution rate. One material consisted of UO2 with added lanthanide oxides (UO2 N) and the other consisted of UO2 with added lanthanide oxides and noble metals (UO2-H) at concentrations to simulate high burnup fuel. The electrochemical corrosion tests on the SIMFUEL materials were conducted in aqueous electrolyte solutions (pH 10) that were either saturated with air or purged with a H2/Ar gas mixture to maintain a constant dissolved H2 concentration. The results from electrochemical corrosion tests on SIMFUEL can be applied to qualitatively understand and also quantify the separate effects of water chemistry and fuel composition on the degradation behavior of the UO2 matrix. Open circuit potential (OCP) measurements on a SIMFUEL material of known composition (i.e., known fraction of NMPs at the fuel surface) immersed in a known solution chemistry (i.e., pH, Eh, [O2], [H2]) at known temperature provide qualitative insight into the degradation behavior of the UO2 matrix. If the OCP is above the threshold potential at which the oxidative dissolution of U(IV) to U(VI) occurs, the SNF is expected to degrade by oxidative processes under the experimental conditions. The net currents that are measured during potentiostatic tests, during which the surface potential of the SIMFUEL material is fixed by a potentiostat, can be used to quantify the UO2 degradation rate and optimize the rate constant values used in the Fuel Matrix Degradation Model (FMDM) for half reactions that occur on the SNF surface. Specifically, electrochemical measurements enable the estimation of the total anodic current at ECORR–which is the surface potential at which the total anodic and total cathodic currents are equal–so that the rate constant values for key reactions can be calculated. The open circuit potential measurements, potentiostatic tests, surface property measurements (scanning electron microscopy images and electrochemical impedence spectroscopy plots), and solution elemental composition analyses are being performed to update and optimize the FMDM. A case study is presented herein to show how electrochemical corrosion test results and accompanying characterization results for the UO2 N material in air-saturated solution can be used to validate the SNF surface reaction module of the FMDM. Future electrochemical tests will be conducted to provide quantitative information on the effects of NM content (burnup), H2 concentration, water chemistry, temperature, and galvanic couples with cladding and EBS alloys on UO2 degradation that can be used to improve the accuracy and functionality of the FMDM.
Follow-on experimentation of irradiated nuclear fuel is critical to understanding fuel behavior at different stages of the fuel lifecycle. Transient experiments using fuel with high burnup can enlarge understanding of fuel fragmentation, relocation and dispersal under various reactor accident situations. Assembling experiments with such fuel requires remote operations.
The AFC-4C irradiation test was irradiated at the Idaho National Laboratory Advanced Test Reactor to investigate several different fast reactor fuels that could be used to facilitate ultra-high burnup applications in sodium fast reactors. Several different alloys, fuel geometries, bonding materials, and coating/barrier were tested in ferritic-martensitic HT-9 cladding. The AFC-4C capsule 5 was removed from the reactor at around 8.7 at %HM burnup to begin postirradiation examination. This report presents and discusses the non-destructive PIE results of the AFC-4C-capsule / rodlet 5, a U-10Zr solid fuel, sodium bonded with HT9 with internal Cr-coating, and its performance is evaluated against the historical fuel performance of previously irradiated fuel from literature.