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This plan describes the post irradiation examination (PIE) oxidation testing activities to be performed by Oak Ridge National Laboratory (ORNL) on individual particles from Compact 5 4 2 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 Plan1 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 4-1-3 taken from the Advanced Gas Reactor (AGR) experiment, AGR 5/6/7. This work will be performed in accordance with the general objectives outlined in the AGR 5/6/7 PIE Plan and guidance in the ORNL PIE Statement of Work.
This report documents the measured post irradiation dimensional change in the AGC-1 samples. The AGC-1 capsule is the first in six planned irradiation capsules comprising the Advanced Graphite Creep (AGC) test series. AGC-1 irradiation began September 5, 2009 in the Advanced Test Reactor (ATR) and was completed on January 8, 2011. The capsule was cooled for 3 months in the ATR Canal, and then shipped to MFC in April 2011 for disassembly and sample extraction. After extraction the samples were shipped to the INL Research Center (IRC) for initial post-irradiation examination (PIE) and storage in the irradiated graphite vault. The AGC-1 capsule design contained “matched pair” samples to ascertain the irradiation-induced dimensional changes and levels of creep experienced in different graphite types. The irradiation-induced dimensional changes and creep levels are determined by comparing the total dimensional change for stressed and unstressed samples of the same type of graphite exposed to the same dose levels and at similar temperatures. Under irradiation creep (i.e. permanent strain due to irradiation, stress, and temperature) the stressed samples should demonstrate more dimensional change than the unstressed samples. This additional dimensional change in the stressed samples is designated as “irradiation-induced creep” in graphite. The data are further presented using the parameters influencing dimensional change in graphite; levels of induced stress, temperature, graphite type, and dose. However, the AGC-1 post-irradiation examination is a significant endeavor and this data report serves to provide irradiation-induced dimensional change data for AGC capsule design refinement as well as a status on the progress of the PIE activities. The dimensional changes of both the samples and graphite body are very important to the design of the future AGC capsules (AGC-3 through AGC-6) and are provided as soon as the data are available in order to determine whether design changes to the next capsule are required. A complete evaluation of the irradiation-induced dimensional change data will be performed for a final AGC-1 PIE report that will include full analysis of pre- and post-irradiation data, with verified AGC-1 irradiation conditions of temperature and dose.
Highlights from the postirradiation examination of the AFC-3A and AFC-3B irradiation test are presented. Baseline non-destructive and destructive examinations of these tests have been completed at this time. Results from visual examination, neutron radiography, gamma spectrometry, dimensional inspection, fission gas release, chemical burnup analysis, and optical microscopy have been collected. All the rodlets from these irradiations appear to have performed well during irradiation. Baseline PIE has been completed on rodlets from AFC- 3A, and AFC-3B. This includes visual exams, neutron radiography, gamma spectrometry. dimensional inspection, fission gas release evaluation, optical microscopy. microhardness evaluation. and chemical analysis. The results compiled here are a sampling of the full results. Baseline PIE provides an engineering or macroscopic scale evaluation of fuel performance and prepares samples for a more detailed phenomenological study of observed performance. Advanced PIE techniques are desired to obtain microscopic elemental and structural information both in the fuel, in the cladding, and in the interface where the composition of the FCCI layer is of great interest. The AFC-3 tests do appear to have performed roughly as designed during the irradiation. The FCCI behavior of Mo alloy metallic fuel versus Zr alloy metallic fuel has been confirmed. Results from this irradiation such as the FCCI observed in AFC-3A R1 and AFC-3A R2 can be used to eliminate Mo alloys as a choice in future irradiations where ultra-high burnup is a stated goal. Annular fuels are a preferred method for many high burnup core designs. The redistribution behavior of the annular U-Zr alloys should be studied further to better understand how this might impact high burnup annular fuel performance. The thermal bonding between annular fuel and cladding should also be studied further as there appear to be localized areas of FCCI forming in both AFC-3A R4 and AFC-3B R2, while no FCCI was observed in the Zr-based Na-bonded or Mo-based low smear density Na-bonded fuel. The performance of the low smear density. solid fuels is less clear. The neutron radiography and optical microscopy cross sections suggest the fuel is not forming a radially symmetric structure which may be causing undesirable localized temperature variations. Further study both by electron microscopy and by simulation to evaluate the implications of the non-symmetric fuel form are needed to better understand whether this fuel form merits further study under irradiation. (authors)
Analytical electron microscopy (AEM) in the scanning transmission electron microscope (STEM) has long been a staple of post-irradiation examination (PIE) and materials development campaigns. Recently, new advances in both STEM hardware and post-processing data-analysis software, along with ancillary advances such as in focused ion beam (FIB) sample preparation, has led to exciting advances in the quantity and quality of data that can be obtained. This talk will summarize recent experiments performed at Oak Ridge National Laboratory (ORNL), especially work on neutron-irradiated materials and fuels taking advantage of a newly-installed latest-generation AEM tool in the ORNL irradiated materials laboratories. ORNL installed a new FEI Talos F200X microscope, a 200 keV, high-brightness Schottky emission AEM system optimized for high-speed, high-efficiency X-ray mapping experiments. The F200X features the FEI 'SuperX' four-sector silicon drift detector (SDD) X-ray detector, which provides 0.9 sr solid angle of collection. By combining the high-brightness electron source, latest-generation electron optics, and fast computers and pulse processors, X-ray mapping speed is increased by a factor of 50-100 x compared to the previous-generation instruments. In addition to improvements in hardware, data mining via multivariate statistical analysis (MVSA) allows for maximum insight to the data obtained from the new hardware. MVSA data mining of the X-ray maps allows, for instance, unbiased computational identification of the phases present, and discovery of small phases or precipitates not expected by the human analyst. This talk will present examples of experiments performed using the most modern AEM tools and software. Silicon carbide is a major candidate for next-generation nuclear applications. Nano-infiltration by transient eutectic (NITE) SiC requires low sintering temperatures. X-ray mapping of the NITE structure find pockets of different oxides, such as Zr-Y-oxide, Y-Al-oxide, and yttria. MVSA data mining finds the regions where different oxides are present and indicates that Y additions preferentially wet grain boundaries, where Zr additions are preferentially concentrated at grain triple points. SiC coatings from AGR-I experiment TRISO particles were found to contain significant quantities of fission products (e.g., Pd) as well as actinides (U, Pu). MVSA data mining found complex precipitates and grain boundary films of the fission products and actinides deep (∼several microns) into the SiC layer. Advanced AEM methods, combined with MVSA data mining, makes possible both rapid and detailed PIE. Ferritic alloys, such as accident-tolerant FeCrAl grades, benefit greatly from advanced AEM examination. Neutron-irradiated FeCrAl precipitates the detrimental Cr-rich α' phase. Because the SuperX SDD system has four detectors, it is possible to obtain high X-ray count rates even on a highly tilted specimen, as might be needed to obtain STEM images of dislocation loops. Results on neutron-irradiated FeCrAl indicates that α' size and number density can be obtained by AEM, and MVSA results determine that minor elements (such as Si) has complex radiation-induced segregation behavior at grain boundaries. We have also developed and validated computer models of the AEM experiment, allowing different microstructures' X-ray map results to be simulated for different STEM instruments. This allows X-ray mapping to be more quantitative, such as by estimating what size nanoclusters in an oxide-dispersion strengthened steel would be visible or invisible in STEM. In summary, new advanced AEM technology provides a major leap forward in experiments possible in PIE of irradiated materials. (authors)
Development of new materials for current nuclear power plants (NPPs) or materials for new NPPs requires extensive research, development, and deployment programs. Central to these efforts, particularly for core internals, are studies on the microstructural and mechanical changes due to neutron irradiation. These radiation effects studies are typically completed using materials test reactors followed by expensive and time-consuming post-irradiation examinations (PIE). This sluggish approach has severely limited the nuclear materials field to develop new materials or verify materials for new reactor design use. It is also very difficult to achieve the high levels of damage proposed for future reactors. A logical approach to accelerating the field is the application of ion beams to simulate neutron damage, especially for high dose applications. Ion beams have been used for decades to study radiation effects but not to emulate high dose reactor irradiations. In this regard, new (or renewed) methods for ion beam irradiations, new developments in materials characterization, and greater understanding on the governing mechanisms controlling observed radiation effects have led to a need for further studies to refine the technique. A primary challenge is the ability to determine the ion irradiation conditions that produce microstructures and mechanical properties relevant to in-core neutron irradiation conditions. A large, multi-disciplinary, multi-laboratory program deemed the High Fidelity Ion Beam Simulation of High Dose Neutron Irradiation has been formed to assess and overcome the challenges associated with emulating high dose neutron irradiations using ion beams. Central to this program is the use of dual-beam ion irradiations and corresponding neutron irradiations within the BOR-60 fast reactor. These irradiations have produced hundreds of specimens for microstructural and mechanical characterization. A significant issue regarding the samples is the radioactivity of the neutron irradiated specimens as well as the need to characterize both irradiation types using the same equipment to maintain consistency and repeatability. The use of the Nuclear Science User Facilities (NSUF) has been central towards overcoming the mentioned issues and thereby enabling robust PIE in both the ion and neutron irradiated specimens of interest. In particular, the electron microscopes housed within the Low Activation Materials Development and Analysis (LAMDA) facility at Oak Ridge National Laboratory (ORNL) - an NSUF partner facility - has been instrumental in gathering high quality data for analysis. This paper will briefly highlight the PIE efforts using the NSUF capabilities within LAMDA.
Post irradiation examination (PIE) of accident tolerant fuel concepts and high burnup uranium dioxide advanced zirconium alloy clad fuel contribute to near term nuclear industry goals of enhancing light water reactor safety and economics. Westinghouse is developing chromium coated zirconium alloy cladding as an accident tolerant fuel option for current pressurized water reactors. Examination of high burnup fuel further augments the technical foundation for extending the peak rod average burnup limits beyond the current regulatory limit of 62 MWd/kgU. Seven fuel rods were received at the Oak Ridge National Laboratory Irradiated Fuel Examination Facility. This includes three chromium coated zirconium alloy fuel rods and four high burnup advanced zirconium alloy fuel rods. Six of the seven rods were fueled with standard uranium dioxide while one of the chromium coated rods contained Westinghouse’s ADOPT fuel. The overall hot-cell PIE plan and some PIE results have been presented previously. This paper focuses on the additional non-destructive examination and fission gas release. Highlights from visual examination of these rods are presented. The chromium coating showed little to no visual change following irradiation. Observations of the axial burnup trends in the fuel rods are derived from full rod axial gamma-ray scans with additional observations related to local fission product migration. The axial dimensional changes of the seven rods are evaluated. As expected, creep down had not yet occurred in the chromium coated rods, and the maximum diameter strain in the high burnup advanced zirconium alloy cladding was less than 0.5%. Finally, the steady state fission gas release for these rods was measured ranging from 4.1% to 17.6% which matched expectation from literature. These fuel rods now await further destructive examination.
Post irradiation examinations (PIE) were performed on irradiated iron-chromium-aluminum (FeCrAl) specimens. These FeCrAl specimens were fabricated at the US Department of Energy's Oak Ridge National Laboratory (ORNL). The experimental setup involved subjecting FeCrAl cladding, along with UO 2 pellets, to irradiation in the Idaho National Laboratory Advanced Test Reactor (ATR). In parallel, the FeCrAl alloy tubing without UO 2 pellets was irradiated at ORNL's High Flux Isotope Reactor (HFIR). After irradiation, the ATR-irradiated rodlet was transported to an ORNL hot cell, where it was sectioned into multiple samples for the PIE and severe-accident testing. The sectioning process revealed that the fuel was not bonded to the cladding and could be easily detached from sectioned cladding slices. Microstructural analysis of the fuel cross sections demonstrated no significant interaction between the fuel and the cladding. Additionally, high-temperature steam oxidation tests on defueled cladding segments showed minimal oxygen uptake even at 1200 °C. Here, the ATR-irradiated specimens began to exhibit signs of enhanced oxidation upon reaching a temperature of 1300 °C. Furthermore, enhanced oxidation was observed on the inner surface of the ATR-irradiated FeCrAl specimen, which had been subjected to 1300 °C for a duration of 1 min. By contrast, high-temperature steam oxidation experiments indicated that the HFIR-irradiated FeCrAl cladding provided good thermal stability when exposed to 1300 °C for up to 4 h. Comparative analysis encompassing the oxidation behavior of the ATR-irradiated fueled FeCrAl, HFIR-irradiated unfueled FeCrAl, and unirradiated FeCrAl suggests that the fuel–cladding interaction, although not visible via standard microscale electron microscopy measurements, may accelerate the deterioration of FeCrAl cladding in beyond-design-basis accident scenarios.
This report documents comparisons between post-irradiation examination measurements and model predictions of silver (Ag), cesium (Cs), and strontium (Sr) release from selected tristructural isotropic (TRISO) fuel particles and compacts during the first irradiation test of the Advanced Gas Reactor program that occurred from December 2006 to November 2009 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). The modeling was performed using the particle fuel model computer code PARFUME (PARticle FUel ModEl) developed at INL. PARFUME is an advanced gas-cooled reactor fuel performance modeling and analysis code (Miller 2009). It has been developed as an integrated mechanistic code that evaluates the thermal, mechanical, and physico-chemical behavior of fuel particles during irradiation to determine the failure probability of a population of fuel particles given the particle-to-particle statistical variations in physical dimensions and material properties that arise from the fuel fabrication process, accounting for all viable mechanisms that can lead to particle failure. The code also determines the diffusion of fission products from the fuel through the particle coating layers, and through the fuel matrix to the coolant boundary. The subsequent release of fission products is calculated at the compact level (release of fission products from the compact) but it can be assessed at the particle level by adjusting the diffusivity in the fuel matrix to very high values. Furthermore, the diffusivity of each layer can be individually set to a high value (typically 10-6 m2/s) to simulate a failed layer with no capability of fission product retention. In this study, the comparison to PIE focused on fission product release and because of the lack of failure in the irradiation, the probability of particle failure was not calculated. During the AGR-1 irradiation campaign, the fuel kernel produced and released fission products, which migrated through the successive layers of the TRISO-coated particle and potentially through the compact matrix. The release of these fission products was measured in PIE and modeled with PARFUME.
The AGR 2 irradiation experiment began in June 2010 and was completed in October 2013. The test train was shipped to the Materials and Fuels Complex in July 2014 for post-irradiation examination (PIE). The first PIE activities included nondestructive examination of the test train, followed by disassembly of the test train and individual capsules and detailed inspection of the capsule contents, including the fuel compacts and their graphite fuel holders. Dimensional metrology was then performed on the compacts, graphite holders, and steel capsule shells. AGR 2 disassembly and metrology were performed with the same equipment used successfully on AGR 1 test train components. Gamma spectrometry of the intact test train gave a preliminary look at the condition of the interior components. No evidence of damage to compacts or graphite components was evident from the isotopic and gross gamma scans. Disassembly of the AGR 2 test train and its capsules was conducted rapidly and efficiently by employing techniques refined during the AGR 1 disassembly campaign. Only one major difficulty was encountered while separating the test train into capsules when thermocouples (of larger diameter than used in AGR 1) and gas lines jammed inside the through tubes of the upper capsules, which required new tooling for extraction. Disassembly of individual capsules was straightforward with only a few minor complications. On the whole, AGR 2 capsule structural components appeared less embrittled than their AGR 1 counterparts. Compacts from AGR 2 Capsules 2, 3, 5, and 6 were in very good condition upon removal. Only relatively minor damage or markings were visible using high resolution photographic inspection. Compact dimensional measurements indicated radial shrinkage between 0.8 to 1.7%, with the greatest shrinkage observed on Capsule 2 compacts that were irradiated at higher temperature. Length shrinkage ranged from 0.1 to 0.9%, with by far the lowest axial shrinkage on Capsule 3 compacts—possibly as a consequence of lower packing fraction or larger particle size. Differences in fast neutron fluence among compacts from these four capsules had no obvious effect on radial and axial shrinkage. (The AGR 2 experiment included Capsule 1 containing French compacts and Capsule 4 with compacts made at Oak Ridge National Laboratory using South African fuel particles. Information on these two batches of AGR 2 fuel compacts is confined to restricted Appendices A and B because of proprietary information limitations.)
Post-irradiation examination (PIE) is being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2). This effort builds upon the understanding acquired throughout the AGR-1 PIE campaign, and is establishing a database for the different AGR-2 fuel designs. The AGR-2 irradiation experiment included TRISO fuel particles coated at BWX Technologies (BWXT) with a 150-mm-diameter engineering-scale coater. Two coating batches were tested in the AGR-2 irradiation experiment. Batch 93085 had 508-μm-diameter uranium dioxide (UO 2 ) kernels. Batch 93073 had 427-μm-diameter UCO kernels, which is a kernel design where some of the uranium oxide is converted to uranium carbide during fabrication to provide a getter for oxygen liberated during fission and limit CO production. Fabrication and property data for the AGR-2 coating batches have been compiled and compared to those for AGR-1. The AGR-2 TRISO coatings were most like the AGR-1 Variant 3 TRISO deposited in the 50-mm-diameter ORNL lab-scale coater. In both cases argon-dilution of the hydrogen and methyltrichlorosilane coating gas mixture employed to deposit the SiC was used to produce a finer-grain, more equiaxed SiC microstructure. In addition to the fact that AGR-1 fuel had smaller, 350-μm-diameter UCO kernels, notable differences in the TRISO particle properties included the pyrocarbon anisotropy, which was slightly higher in the particles coated in the engineering-scale coater, and the exposed kernel defect fraction, which was higher for AGR-2 fuel due to the detected presence of particles with impact damage introduced during TRISO particle handling.
The PARFUME (PARticle FUel ModEl) code was used to predict silver release from tristructural isotropic (TRISO) coated fuel particles and compacts during the second irradiation experiment (AGR-2) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-2 experiment used the fuel compact volume average temperature for each of the 559 days of irradiation to calculate the release of fission product silver from a representative particle for a select number of AGR-2 compacts and individual fuel particles containing either mixed uranium carbide/oxide (UCO) or 100% uranium dioxide (UO2) kernels. Post-irradiation examination (PIE) measurements were performed to provide data on release of silver from these compacts and individual fuel particles. The available experimental fractional releases of silver were compared to their corresponding PARFUME predictions. Preliminary comparisons show that PARFUME under-predicts the PIE results in UCO compacts and is in reasonable agreement with experimental data for UO2 compacts. The accuracy of PARFUME predictions is impacted by the code limitations in the modeling of the temporal and spatial distributions of the temperature across the compacts. Nevertheless, the comparisons on silver release lie within the same order of magnitude.
This plan describes post-irradiation examination (PIE) activities (safety testing and post-safety test analyses) to be performed by Oak Ridge National Laboratory (ORNL) on irradiated Compact 4-1-2 taken from the Advanced Gas Reactor (AGR) experiment, AGR 2. This work will be performed in accordance with the Statement of Work for PIE and Safety Testing of Pebble Bed Modular Reactor Limited (PBMR) Fuel at ORNL.
Advancing the deployment of sodium-cooled fast reactors (SFRs) requires thorough testing of metallic fuel pins under accident conditions to establish safe operational limits of high burnup fuel. To conduct transient testing, a comprehensive understanding of steady-state fuel behavior obtained through both experimental characterization and accurate predictive capabilities is needed. This study comparatively assesses the steady-state irradiation performance of two high burnup U-19Pu-10Zr fuel pins, DP-36 and DP-40, irradiated under prototypic fast reactor conditions in preparation for planned safety testing at the Transient Reactor Test Facility. Since DP-40 was designated for use in the test and DP-36 serves as its sibling pin, non-destructive, engineering-scale post-irradiation examinations (PIE) were conducted on both pins while destructive examinations were performed exclusively on DP-36. The results were then assessed against historical performance data from similar fuel pins irradiated in the Experimental Breeder Reactor-II. Additionally, the steady-state irradiation of each pin was modeled using the BISON fuel performance code to assess the accuracy of current modeling capabilities in predicting the baseline irradiation behavior. Non-destructive examinations included neutron radiography to measure fuel column elongation, gamma scanning to verify pin integrity and fission product migration, and profilometry to assess dimensional changes. Benchmarking against existing PIE data revealed consistent patterns in axial fuel column growth and cladding diametral strain, though both pins exhibited longer low-density “fluff” structures, which can have implications for core reactivity and source term calculations. Destructive examinations on DP-36 included fission gas release analysis and sectioning for optical microscopy, which showed more complex constituent redistribution patterns than the traditionally accepted 3-ring model. The axial evolution of fractional areas and porosities of each of the redistributed zones were quantified and presented. Modeling comparisons showed agreement in fractional fission gas release but consistently overestimated axial and radial swelling and disagreed with measured axial porosity patterns. These conservative overpredictions suggested that the pins would appear closer to failure or operational limits at the start of transient tests, potentially leading to higher strain accumulation during the transient. While conservative estimates provide safety margins, they can negatively impact fuel economics. A review of the swelling models identified areas for improvement in the gaseous swelling, solid swelling, and fuel hot-pressing models when applied to ternary fuel. The results of this study highlight the critical importance of conducting pre-test characterization on both test and sibling pins to accurately capture steady-state fuel behavior, providing a precise baseline for post-test evaluations and essential inputs for transient modeling of the planned experiments. The analysis also revealed significant data gaps that require further investigation to enhance the understanding and prediction of fuel swelling and pore dynamics. Collecting comprehensive data across different irradiation conditions, burnup levels, and fuel compositions are essential for refining existing models and developing mechanistic models for both binary and ternary metallic fuels, ultimately improving the integration of modeling and experimental approaches in accident testing.
The main purpose of this study is the modification of the parametric integral equations systems (PIES) method, to include the NURBS curves into its mathematical formulas. We want to create an opportunity for defining the boundary shape using NURBS curves without the classical discretization process. We present an inclusion of the NURBS curves into PIES, and its application in process of modeling and solving the boundary value problems, on Laplace’s equation example. The correctness and accuracy of such modeling is confirmed by the examples. Obtained solutions are compared with these obtained using linear segments and B-spline curves. Additionally, we verify the correctness of modeling and modifying the shape using analytical solutions. The NURBS curves significantly expand the possibilities and increase the accuracy of the boundary shape modeling which is an important aspect of the numerical solving of boundary value problems.
The control of the interaction between the U(Mo) fuel phase and the Al matrix is one of the challenges of dispersion fuel plate development for research reactors. Given the specific properties of this interaction layer, larger amounts of it in the meat could lead to a reduction of the plate mechanical integrity and thermal conductivity, eventually leading to pillowing. The SELENIUM project showed that by depositing a ZrN coating on the surface of the U(Mo) fuel particles, the amount of formed U(Mo)-Al interaction layer is limited but still present. Microstructural analysis performed on the as fabricated coating and fresh fuel plates containing ZrN coated U(Mo) dispersed in an Al matrix, revealed that the coating gets damaged during plate production. The post irradiation examinations (PIE) of the ZrN coated U(Mo) fuel plates, from the SELENIUM and SEMPER FIDELIS experiments, show how the U(Mo)-Al interaction layer is formed -only at those locations where the coating is missing or damaged -and the evolution of coating microstructure during irradiation. Finally, as a remedy, to further reduce the amount of interaction layer formed, the use of an Al-Si matrix was proposed based on the higher affinity of Si for U compared to the affinity of Al for U. PIE of a fuel plate consisting of ZrN coated U(Mo) dispersed in an Al-Si matrix irradiated in the SEMPER FIDELIS experiment, clearly demonstrates the benefit of adding Si to the matrix.
Post-irradiation examination (PIE) is critical for the performance assessment and qualification of nuclear fuels. Secondary ion mass spectrometry (SIMS) is a powerful materials characterization technique that allows for elemental and isotopic mapping with a depth resolution greater than EDS and EPMA. However, it has not yet been applied to PIE of metallic nuclear fuel. Here, in this work, we characterize an fast neutron spectrum irradiated U-10Zr fuel sample using a time-of-flight SIMS (ToF-SIMS) system connected to a FIB/SEM system, which allows for flexible sample analysis compared to a dedicated ToF-SIMS instrument. Analysis of the resulting hyperspectral micrograph data was aided by the development of an unsupervised machine learning (ML) algorithm that iterates on existing methods to segment the 3D micrographic datasets based on the similarity of mass spectra. The results showed that the FIB-ToF-SIMS instrument was potentially capable of spatially resolving closed fission gas bubbles in 3D by continued ion sputtering of the analyzed volume. Additionally, the ML algorithm proved useful in revealing the chemical segregation of light fission products (those with an atomic mass between approximately 85–105 amu, such as ruthenium and rhodium) plus matrix zirconium, heavy fission products (those with an atomic mass between approximately 135–150 amu, such as the lanthanides) and uranium. Future studies are planned to conduct FIB-ToF-SIMS analysis on more irradiated U-Zr samples to study the constituent redistribution.
As part of the High Temperature Reactors (HTR) R&D program, a series of irradiation tests, designated as Advanced Gas-cooled Reactor (AGR), have been defined to support development and qualification of fuel design, fabrication process, and fuel performance under normal operation and accident conditions. The AGR tests employ fuel compacts placed in a graphite cylinder shrouded by a steel capsule and instrumented with thermocouples (TC) embedded in graphite blocks enabling temperature control. While not possible to obtain by direct measurements in the tests, crucial fuel conditions (e.g., temperature, neutron fast fluence, and burnup) are calculated using core physics and thermal modeling codes. This paper is focused on AGR test fuel temperature predicted by the ABAQUS code's finite element-based thermal models. The work follows up on a previous study, in which several statistical analysis methods were adapted, implemented in the NGNP Data Management and Analysis System (NDMAS), and applied for qualification of AGR-1 thermocouple data. Abnormal trends in measured data revealed by the statistical analysis are traced to either measuring instrument deterioration or physical mechanisms in capsules that may have shifted the system thermal response. The main thrust of this work is to exploit the variety of data obtained in irradiation and post-irradiation examination (PIE) for assessment of modeling assumptions. As an example, the uneven reduction of the control gas gap in Capsule 5 found in the capsule metrology measurements in PIE helps identify mechanisms other than TC drift causing the decrease in TC readings. This suggests a more physics-based modification of the thermal model that leads to a better fit with experimental data, thus reducing model uncertainty and increasing confidence in the calculated fuel temperatures of the AGR-1 test.