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This plan describes post-irradiation examination (PIE) activities to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 3-4-1 that was taken from the Advanced Gas Reactor (AGR) experiment, AGR-2. This work will be performed in accordance with general objectives outlined in the AGR-2 PIE plan and guidance in the ORNL PIE statement of work.
This plan describes post-irradiation examination (PIE) activities (simultaneous three-compact temperature transient safety testing and optional post-safety test analyses) to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compacts 5-1-1, 5-1-2, and 5-3-1 taken from the Advanced Gas Reactor (AGR) experiment, AGR 2. This work will be performed in accordance with general objectives outlined in the AGR-2 PIE plan and guidance in the ORNL PIE statement of work.
This plan describes the post irradiation examination (PIE) activities to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 6-2-3 taken from the Advanced Gas Reactor (AGR) experiment, AGR 2. This work will be performed in accordance with the general objectives outlined in the AGR 2 PIE Plan and guidance in the ORNL PIE Statement of Work.
This plan describes the post irradiation examination (PIE) activities to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 2-2-1 taken from the Advanced Gas Reactor (AGR) experiment, AGR 2. This work will be performed in accordance with the general objectives outlined in the AGR 2 PIE Plan and guidance in the ORNL PIE Statement of Work.
This plan describes the post irradiation examination (PIE) activities to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 7 4 taken from the Advanced Gas Reactor (AGR) experiment, AGR 3/4. This work will be performed in accordance with the general objectives outlined in the AGR 3/4 PIE Plan and guidance in the ORNL PIE Statement of Work.
This plan describes the post irradiation examination (PIE) activities to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 8 4 taken from the Advanced Gas Reactor (AGR) experiment, AGR 3/4. This work will be performed in accordance with the general objectives outlined in the AGR 3/4 PIE Plan and guidance in the ORNL PIE Statement of Work.
Post-irradiation examination (PIE) is being carried out on coated particle fuel compacts from the first Advanced Gas Reactor (AGR) Fuel Development and Qualification irradiation experiment (AGR-1). The Oak Ridge National Laboratory (ORNL) Core Conduction Cooldown Test Facility (CCCTF) has been used to evaluate the effect of elevated temperature on the fuel microstructure and fission product retention by heating compacts to maximum temperatures of 1600, 1700, or 1800°C for typically 300 hours, where 1600°C is the expected maximum temperature during a high-temperature gas-cooled reactor (HTGR) depressurization conduction cooldown event, while 1700 and 1800°C explore the performance margin. After completing eleven standard safety tests, a different type of experiment was conducted [INL/PLN4650], where, rather than heating an entire as-irradiated fuel compact, loose particles were tested after deconsolidation of a compact and separation of the particles from the compact matrix.
The third Advanced Gas Reactor (AGR) irradiation experiment, AGR-3/4, was designed to investigate the migration of fission products in fuel compact graphitic matrix and reactor graphite components. Using destructive methods, radial fission product concentration profiles were measured for gamma-emitting fission products (e.g., Ag-110m, Cs-134, and Eu-154) and beta-emitting Sr-90 in irradiated graphitic and graphite components from six different AGR-3/4 irradiation capsules. These new measured concentration profiles can now be compared to non-destructive measurements and fission product transport simulations and will be used to derive new diffusivities and sorptivities to support refinement of fission product transport models and high-temperature gas-cooled reactor (HTGR) source-term analyses. Each capsule in the AGR-3/4 experiment had four fuel compacts in the middle of two concentric rings of graphitic matrix material, PCEA graphite, or IG-110 graphite. In addition to the approximately 1898 tristructural isotropic (TRISO) coated particles in each compact, there were 20 designed to fail (DTF) particles coated only in pyrocarbon so that they released fission products into the surrounding cylindrical rings of carbonaceous materials. Destructive sampling of the rings involved machining/milling material from around the circumference of the rings, collecting that material, and performing radiochemical analyses on it. Milling operations were performed in multiple steps or segments, and each segment was generally 0.508 mm (0.020 in) thick. Knowing the radial position at which each segment was milled, the volume of the milled material at each segment, and the fission product content in each segment, the radial fission product concentrations were constructed for select isotopes in each ring. Ag-110m profiles had the most variation. Some profiles were peaked at an inner or outer surface. Some were peaked at the middle of the ring wall thickness. Some increased radially outward, and some decreased radially outward. These types of variations and the fact that the measured profiles do not generally compare favorably with the transport model employed for AGR-3/4 may adversely impact the ability to extract reasonable transport parameters for this isotope. In many cases, the Cs-134 profiles decreased somewhat linearly in the outward radial direction, and in cursory comparisons, the shapes of these profiles appeared similar to those from model predictions. The step changes in concentration across the inner-outer ring gap were generally consistent with the model predictions as well. In some cases, there were local maxima in concentration at the outer surface of the rings. This suggests that fission products could have transported in the small gaps between the inner ring and the outer ring and between the outer ring and the sink ring such that some portion of a given fission product can bypass diffusion through the ring itself. The analysis of the small nubs on the outer surfaces of some of the outer rings revealed fission product concentrations in the nubs that were often higher than in the outermost segments of the rings. This further supports the hypothesis that short-circuit, gap transport occurred, causing relatively high surface concentrations on the outer surfaces of the rings. Eu-154 and Sr-90 profiles tend to have very similar shapes, suggesting that they transport via the same mechanisms. The observed profiles were indicative of a transport process where the isotopes are sorbed on the inner surface of the ring, but diffusion into the ring from that surface is quite slow. Some elevated concentrations of Sr-90 (relative to Eu-154) on the outer surface of a ring suggested that rapid, gas-gap transport of gaseous precursor Kr-90 and volatile Rb-90 could have occurred prior to their decaying to Sr-90. Overall, the Eu 154 and Sr-90 profiles were still very similar, which indicates that the transport of short-lived Sr-90 precursors is not a major effect. In some capsules, the qualitative Sr-90 behavior across the ring gaps was consistent with the model (using the available legacy Sr-90 transport parameters), but in other capsules the model was inconsistent with the measurements and seems to underestimate the amount of Sr-90 in the outer rings. The total ring Sr-90 inventories were estimated for all the rings that were subject to physical sampling. These results will be used to adjust the predicted particle and/or compact releases used in the AGR-3/4 fission product transport model. Given the different irradiation temperatures among the capsules and the rings, it was not possible to discern fundamental differences in the transport of isotopes within the different carbon materials, i.e., graphitic matrix, IG-110, or PCEA. It may be possible to do this in the course of determining transport from the concentration profiles in future work.
This report summarizes the activities performed in FY-25 to support fuel performance modeling for the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program. This includes implementation of a new uranium oxycarbide (UCO) kernel swelling rate model, inner pyrolytic carbon (IPyC) cracking behavior and failure predictions, development of new UCO and silicon carbide (SiC) cesium diffusion parameters, and the thermomechanical behavior of particle layers using experimental data from micro-tensile strength testing.
Post irradiation examination (PIE) and heating tests of tristructural isotropic (TRISO)-coated fuel particles, fuel compacts, and graphite from Advanced Gas Reactor (AGR) irradiation experiment AGR 3/4 are in progress. The AGR-3/4 test train consisted of twelve separate capsules irradiated in the northeast flux trap of the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) from December 14, 2011, to April 12, 2014 (Collin 2015). The plan for the reirradiation in the Neutron Radiography Reactor (NRAD) and subsequent 1200°C heating test of AGR-3/4 Compact 8-1 in the Fuel Accident Condition Simulator (FACS) furnace was described previously in (Stempien 2019). One of the main goals of that test was to assess the retention of short-lived fission products I-131 (t1/2 = 8.02 d) and Xe-133 (t1/2 = 5.2 d) in fuel kernels. Xe-133 release had ostensibly leveled off after approximately 80 hours. An issue with the condensation plate exchanges forced this test to be stopped after approximately 96 hours. The issue was resolved, and the test was restarted by heating FACS back up to 1200°C. During this reheating phase, an additional release of Xe-133, approximately equal in size to the release during the initial heating to 1200°C, was detected in the fission gas monitoring system (FGMS). To test whether this release is related to temperature transients (i.e. from heating and reheating), this new plan will describe three additional temperature cycles.
This paper describes the concept and preliminary component testing of a gas-cooled, UN-fueled, pin-type reactor which uses He/Xe gas that goes directly into a recuperated Brayton system to produce electricity for nuclear electric propulsion. This Direct-Drive Gas-Cooled Reactor (DDG) is designed to be subcritical under water or wet-sand immersion in case of a launch accident. Because the gas-cooled reactor can directly drive the Brayton turbomachinery, it is possible to configure the system such that there are no external surfaces or pressure boundaries that are refractory metal, even though the gas delivered to the turbine is 1144 K. The He/Xe gas mixture is a good heat transport medium when flowing, and a good insulator when stagnant. Judicious use of stagnant cavities as insulating regions allows transport of the 1144-K gas while keeping all external surfaces below 900 K. At this temperature super-alloys (Hastelloy or Inconel) can be used instead of refractory metals. Super-alloys reduce the technology risk because they are easier to fabricate than refractory metals, we have a much more extensive knowledge base on their characteristics, and, because they have a greater resistance to oxidation, system testing is eased. The system is also relatively simple in its design: no additional coolant pumps, heat exchanger, or freeze-thaw systems are required. Key to success of this concept is a good knowledge of the heat transfer between the fuel pins and the gas, as well as the pressure drop through the system. This paper describes preliminary testing to obtain this key information, as well as experience in demonstrating electrical thermal simulation of reactor components and concepts.
The purpose of this study is to identify the material properties that have the largest impact on the failure probability of tristructural isotropic (TRISO)-coated fuel particles under irradiation. For the most part, these material properties were obtained from experimental data of historical TRISO fuel development programs in Europe and the United States. The adequacy of these historical constitutive material properties is being evaluated as part of the Advanced Gas Reactor Fuel Development and Qualification Program. The TRISO fuel performance modeling code PARFUME was used to assess the material properties that have the main influence on the probability of failure of the silicon carbide (SiC) layer of TRISO fuel particles. Although some of these material properties exhibit a large variability according to published data, only a few of them have a significant impact on the SiC failure probability, namely the irradiation-induced creep and dimensional change of the inner pyrolytic carbon layer.
A variety of neutron irradiation experiments were completed at Idaho National Laboratory (INL) on tristructural isotropic (TRISO) coated particles contained in a graphitic matrix as part of the Advanced Gas Reactor (AGR) fuel development and qualification program. Although the initial advanced microscopy and microanalysis studies on AGR-1 particles focused predominantly on the SiC layer?s role as the main fission-product containment, studies were further expanded to evaluate UCO kernels of selected AGR-1 and AGR-2 coated particles. Fuel kernels tested in the AGR-1 and AGR-2 experiments is composed of a mixture of UO2 and UC, with a nominal as-fabricated stoichiometry of UO1.37 C0.35 and UO1.43 C0.39, respectively. These as-fabricated kernel-stoichiometry values are determined at the completion of kernel fabrication and do not account for any changes that will occur during the subsequent coating and compacting processes.
Graphitic matrix material encapsulates tristructural isotropic (TRISO) coated fuel particles and is a structural component for both pebble and cylindrical fuel compacts in high temperature gas reactor (HTGR) designs. In an off-normal event involving steam exposure, the matrix material is exposed to oxidants, leading to degradation of the fuel compact and subsequent exposure of the TRISO particle fuel to oxidizing species. This study presents microstructural characterization of matrix material after exposure to high temperature (1200°C), 48 kPa steam in which the evolved microstructure demonstrated nonuniform degradation. Subsequent Raman spectroscopy determined the nature of the remaining material post-exposure. Finally, the electron microscopy characterization suggests and Raman spectroscopic analysis confirms, under off-normal conditions, the carbonized phenolic resin binder is preferentially oxidized ahead of the graphite flake filler and is responsible for the nonuniform degradation and enhanced depth of attack.
The Advanced Gas Reactor (AGR)-3/4 irradiation experiment is the third in a series of test irradiations performed as part of the AGR Fuel Development and Qualification Program. AGR-3/4 was designed to investigate the migration of fission products in fuel kernels, compact matrix, and reactor graphite components. The experiment consists of fuel compacts containing driver fuel particles and special “designed-to-fail” particles designed to release fission products during the irradiation, which will migrate through the surrounding cylindrical rings of compact matrix and nuclear grade graphite. Evaluation of the fission product distribution in these rings following the irradiation is a critical component of the experiment, providing data that will support refinement of fission product transport models. Following nondestructive gamma scanning of the inner and outer rings from the standard capsules and any disassembled fuel bodies (such as Capsule 4), physical samples will be taken from selected rings for subsequent analysis for fission products. This activity will provide quantitative data on the inventory and distribution of fission products at specific locations in the rings to compare with the gamma scanning data and model predictions.
This report provides an initial study for producing reduced-order models (ROMs) of pebble-bed high temperature gas reactor (HTGR) models for the purposes of design optimization. As an initial study, this work is meant to be exploratory---identifying useful workflows and methods for ROM generation---and not meant to be a catch-all analysis of HTGR ROM generation and usage for optimization. This report summarizes three tasks performed in Fiscal Year 2022: 1) the creation of HTGR model, 2) the sensitivity analysis of model design parameters, and 3) an introduction to ROM generation techniques. The representative HTGR model created in this work is a multiphysics equilibrium-core using the BlueCRAB (comprehensive reactor analysis bundle) reactor analysis application, coupling four physical phenomena: neutronics, streamline depletion, porous flow thermal hydraulics, and pebble heat conduction. Part of the model creation was identifying some design parameters and quantities of interest that are relevant in an optimization analysis and adjustable in the model. The sensitivity analysis utilized a polynomial chaos expansion methodology to compute global sensitivity metrics. This analysis showed that thermal hydraulics parameters and quantities of interest had a relatively small impact on simulation results. Finally, the ROM generation work involved exploring three different ROM methodologies: polynomial regression, a Gaussian process, and artificial neural networks. Using a cross-validation technique to characterize ROM performance, the Gaussian process and single-layer artificial neural networks showed the most promising results. Overall, this study was insightful and the lessons learned will be invaluable for the eventual development of an HTGR design optimization workflow.
A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN-3636). The AGR-2 test was inserted in the B-12 position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in June 2010 and successfully completed irradiation in October 2013, resulting in irradiation of the TRISO fuel for 559.2 effective full power days (EFPDs) during approximately 3.3 calendar years. The AGR-2 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS) (Pham and Einerson 2014). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as-run thermal analysis has been performed separately on each of four AGR-2 U.S. capsules for the entire irradiation as discussed in (Hawkes 2014). The ABAQUS code’s finite element-based thermal model predicts the daily average volume-average fuel temperature and peak fuel temperature in each capsule. This thermal model involves complex physical mechanisms (e.g., graphite holder and fuel compact shrinkage) and properties (e.g., conductivity and density). Therefore, the thermal model predictions are affected by uncertainty in input parameters and by incomplete knowledge of the underlying physics leading to modeling assumptions. Therefore, alongside with the deterministic predictions from a set of input thermal conditions, information about prediction uncertainty is instrumental for the ART program decision-making. Well defined and reduced uncertainty in model predictions helps increase the quality of and confidence in the AGR technical findings.
To evaluate the stability of nuclear materials in high temperature gas reactors under air ingress conditions, catalytic oxidation of IG-110 graphite by two simulated fission products, metallic Pd and Ag, was studied in oxidative atmosphere and at temperatures up to 1000°C using an integrated furnace, mass spectroscopy and infrared spectroscopy system. Transmission electron microscopy and X-ray diffraction studies show that Pd and Ag nanoparticles were successfully introduced onto powdery IG-110 graphite through an impregnation and subsequent heat-treatment process. The combined mass spectroscopy and infrared spectroscopy methods allow simultaneous analysis of two gaseous products, CO and CO 2 , and separate measurements of activation energy for their formation reactions. It was found that the introduction of Pd or Ag to IG-110 graphite substantially catalyzed the oxidation of graphite, characteristic of decreased onset temperatures for the oxidation of graphite. Moreover, the catalytic effects by Pd and Ag are considerably different based on measured concentration ratios of CO 2 to CO as a function of oxidation temperatures. Ag makes the graphite oxidation commence at approximately 400°C with CO 2 being the dominant product. In contrast, Pd significantly increases the concentration ratio of CO 2 to CO at temperatures higher than approximately 690°C, although it decreases the onset temperature for the oxidation reaction to around 525°C. To understand the catalytic difference, the mechanism of the graphite oxidation was discussed based on the changes of surface oxygen species on Ag and Pd.