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At least 289 records · Page 16

Measurement of kernel swelling and buffer densification in irradiated UCO and UO 2 TRISO fuel particles from AGR-2

The second Advanced Gas Reactor irradiation (AGR-2) featured UCO tristructural isotropic (TRISO)-coated particle fuel and, for comparison purposes, UO2 TRISO fuel. Particles from three UCO fuel compacts and one UO2 fuel compact were chosen for analysis of kernel swelling and buffer shrinkage. Both kernel swelling and buffer shrinkage are common elements of TRISO fuel performance models. More recently, the AGR program has determined that buffer shrinkage may impact the integrity of the inner pyrolytic carbon (IPyC) layer, which may impact SiC layer integrity (particularly during high-temperature heating tests). Therefore, measurements of buffer irradiation-induced shrinkage may be useful in improving existing models and developing new models that capture buffer-IPyC-SiC-fission product interactions observed in recent AGR post-irradiation examination (PIE). TRISO particles were mounted in epoxy, ground, polished, and imaged via optical microscopy in four separate iterations. Each of the four iterations revealed progressively deeper cross sections within the particles. Images collected from each iteration were analyzed for the circumferences/radii of the kernel, buffer, IPyC, and SiC. Spheres were fit to each set of measurements to generate spherical radii representative of the components of the fuel particle. From these radii, volumes were computed and compared to the as-fabricated volumes. UCO kernel swelling was similar among the three AGR-2 compacts ranging from approximately 28 to 32%. This is only slightly higher swelling than was measured in an earlier study of an AGR-1 compact. However, the burnup at which this swelling occurred in AGR-2 was lower than in AGR-1, indicating a faster swelling rate for the AGR-2 kernels for the irradiation conditions covered by these compacts. The AGR-2 UO2 kernel swelling was less, at 9.9%, comparable to other studies of UO2 kernel swelling. While AGR-2 UCO and UO2 kernels demonstrated different rates of swelling, the extent of buffer shrinkage was similar among the two fuel types, ranging from 24-28%. For comparison, a prior study showed that AGR-1 Compact 1-3-1 experienced 39% buffer shrinkage. A commonly-observed post-irradiation particle morphology is where the buffer pulls away from the IPyC (at least in some regions of the particle). Besides the possibility of the buffer degrading the IPyC when it pulls away from the IPyC, this gap is significant because it may also reduce heat transfer within the particle. Average AGR-2 buffer-IPyC gaps were similar for the UCO and UO2 particles, ranging from 22.2 to 26.0 µm. This is similar to what was measured for AGR-1 Compact 1-3-1.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Baseline Postirradiation Examination of the FUTURIX-FTA Experiments

Postirradiation Examination (PIE) results from the baseline examination of the FUTURIX-FTA irradiation experiment are presented in this report. The FUTURIX-FTA irradiation was designed to study the fuel performance of candidate transmutation fuels. The FUTURIX-FTA experiment was conducted in the true fast neutron spectrum conditions of the Phénix sodium fast test reactor in France. FUTURIX-FTA contained several compositions that were also irradiated as part of the AFC-1 test in cadmium filtered positions in the Idaho National Laboratory (INL) Advanced Test Reactor (ATR) that approximate the neutron spectrum of a fast reactor. This experiment was a joint collaboration between the Department of Energy (DOE) in the U.S. and the Commissariat á l'Energie Atomique et aux Energies Alternatives (CEA) in France. A joint agreement between DOE and CEA was established in 2004 to conduct an irradiation experiment of metallic and nitride fuels in the Phénix reactor at CEA. Both the FUTURIX-FTA tests and the AFC-1 tests were conducted for the DOE Advanced Fuels Campaign (AFC) that is part of Fuel Cycle Research & Development (FCRD) program. These programs seek to develop and demonstrate the technologies needed to transmute long-lived transuranic actinide isotopes contained in spent nuclear fuel via fast reactor technology. The long term goal of this work is to reduce the radiotoxicity of future high level waste repositories. Post-irradiation examination of irradiated fuels experiments provides data related to in reactor fuel performance and input into future fuel design choices. The compositions of the FUTURIX-FTA pins were designed to test if minor actinides could be incorporated into the fuel of a fast reactor without significantly changing fuel performance observed for previously well studied fuels. The fuel performance of metallic fuel in fast reactors has been well documented. Likewise the irradiation performance of nitride fuels in fast reactors has also been documented. However the existing database for nitride fuels is much smaller than the database for metal fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Post-irradiation Examination and Fission Product Inventory Analysis of AGR-1 Irradiation Capsules

The AGR-1 experiment was the first in a series of Advanced Gas Reactor (AGR) experiments designed to test TRISO fuel under High Temperature Gas Reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post-irradiation examination (PIE) at INL’s Materials and Fuels Complex (MFC). The inventory and distribution of fission products, especially Ag-110m, was assessed and analyzed for all the components of the AGR-1 capsules. This data should help inform the study of fission product migration in coated particle fuel. Gamma spectrometry was used to measure the activity of various different fission products in the different components of the AGR-1 test train. Each capsule contained: 12 fuel compacts, a graphite holder that kept the fuel compacts in place, graphite spacers that were above and below the graphite holders and fuel compacts, gas lines through which a helium neon gas mixture flowed in and out of each capsule, and the stainless steel shell that contained the experiment. Gamma spectrometry results and the experimental techniques used to capture these results will be presented for all the capsule components. The components were assayed to determine the total activity of different fission products present in or on them. These totals are compared to the total expected activity of a particular fission product in the capsule based on predictions from physics simulation. Based on this metric, a significant fraction of the Ag-110m was detected outside the fuel compacts, but the amount varied highly between the 6 capsules. Very small fractions of Cs-137 (<2E-5), Cs-134 (<1e-5), and Eu-154 (<4e-4) were detected outside of the fuel compacts. Additionally, the distribution of select fission products in some of the components including the fuel compacts and the graphite holders were measured and will be discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Advanced Test Reactor National Scientific User Facility: Addressing advanced nuclear materials research

The Advanced Test Reactor National Scientific User Facility (ATR NSUF), based at the Idaho National Laboratory in the United States, is supporting Department of Energy and industry research efforts to ensure the properties of materials in light water reactors are well understood. The ATR NSUF is providing this support through three main efforts: establishing unique infrastructure necessary to conduct research on highly radioactive materials, conducting research in conjunction with industry partners on life extension relevant topics, and providing training courses to encourage more U.S. researchers to understand and address LWR materials issues. In 2010 and 2011, several advanced instruments with capability focused on resolving nuclear material performance issues through analysis on the micro (10-6 m) to atomic (10-10 m) scales were installed primarily at the Center for Advanced Energy Studies (CAES) in Idaho Falls, Idaho. These instruments included a local electrode atom probe (LEAP), a field-emission gun scanning transmission electron microscope (FEG-STEM), a focused ion beam (FIB) system, a Raman spectrometer, and an nanoindentor/atomic force microscope. Ongoing capability enhancements intended to support industry efforts include completion of two shielded, irradiation assisted stress corrosion cracking (IASCC) test loops, the first of which will come online in early calendar year 2013, a pressurized and controlled chemistry water loop for the ATR center flux trap, and a dedicated facility intended to house post irradiation examination equipment. In addition to capability enhancements at the main site in Idaho, the ATR NSUF also welcomed two new partner facilities in 2011 and two new partner facilities in 2012; the Oak Ridge National Laboratory, High Flux Isotope Reactor (HFIR) and associated hot cells and the University California Berkeley capabilities in irradiated materials analysis were added in 2011. In 2012, Purdue University’s Interaction of Materials with Particles and Components Testing (IMPACT) facility and the Pacific Northwest Nuclear Laboratory (PNNL) Radiochemistry Processing Laboratory (RPL) and PIE facilities were added. The ATR NSUF annually hosts a weeklong event called User’s Week in which students and faculty from universities as well as other interested parties from regulatory agencies or industry convene in Idaho Falls, Idaho to see presentations from ATR NSUF staff as well as select researchers from the materials research field. User’s week provides an overview of current materials research topics of interest and an opportunity for young researchers to understand the process of performing work through ATR NSUF. Additionally, to increase the number of researchers engaged in LWR materials issues, a series of workshops are in progress to introduce research staff to stress corrosion cracking, zirconium alloy degradation, and uranium dioxide degradation during in-reactor use.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Characterization of irradiated AISI 316L stainless steel disks removed from the Spallation Neutron Source

Irradiated AISI 316L stainless steel disks were removed from the Spallation Neutron Source (SNS) for post-irradiation examination (PIE) to assess mechanical property changes due to radiation damage and erosion of the target vessel. Topics reviewed include high-resolution photography of the disk specimens, cleaning to remove mercury (Hg) residue and surface oxides, profile mapping of cavitation pits using high frequency ultrasonic testing (UT), high-resolution surface replication, and machining of test specimens using wire electrical discharge machining (EDM), tensile testing, Rockwell Superficial hardness testing, Vickers microhardness testing, scanning electron microscopy (SEM), and energy dispersive spectroscopy (EDS). The effectiveness of the cleaning procedure was evident in the pre- and post-cleaning photography and permitted accurate placement of the test specimens on the disks. Due to the limited amount of material available and the unique geometry of the disks, machine fixturing and test specimen design were critical aspects of this work. Multiple designs were considered and refined during mock-up test runs on unirradiated disks. The techniques used to successfully machine and test the various specimens will be presented along with a summary of important findings from the laboratory examinations.

cavitation-induced erosion↗

US Progress on Property Characterization to Support LEU U-10 Mo Monolithic Fuel Development

The US High Performance Research Reactor program is pursuing development and qualification of a new high density monolithic LEU fuel to facilitate conversion of five higher power research reactors located in the US (ATR, HFIR, NBSR, MIT and MURR). In order to support fabrication development and fuel performance evaluations, new testing capabilities are being developed to evaluate the properties of fuel specimens. Residual stress and fuel-cladding bond strength are two characteristics related to fuel performance that are being investigated. In this overview, new measurement capabilities being developed to assess these characteristics in both fresh and irradiated fuel are described. Progress on fresh fuel testing is summarized and on-going hot-cell implementation efforts to support future PIE campaigns are detailed. It is anticipated that benchmarking of as-fabricated fuel characteristics will be critical to establishing technical bases for specifications that optimize fuel fabrication and ensure acceptable in-reactor fuel performance.

Characterization↗

Test Plan for the 1600°C Heating Test of AGR-3/4 Compact 3-2

Post irradiation examination (PIE) and heating tests of tristructural isotropic (TRISO)-coated fuel particles, fuel compacts, and graphite from the Advanced Gas Reactor (AGR) irradiation experiment AGR 3/4 is 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). This test plan describes the heating test of AGR-3/4 Compact 3-2. The heating test will involve heating the compact in the Fuel Accident Condition Simulator (FACS) furnace to a peak temperature of 1600°C while collecting fission products released from the compact for subsequent analysis. One of several features of AGR-3/4 that set it apart from AGR-1 and AGR-2 is the incorporation of 20 designed-to-fail (DTF) particles in each compact. The fuel kernels in the DTF particles were coated only with a thin (20 µm thick) pyrocarbon layer having high anisotropy such that the coating should most-likely fail during the irradiation, resulting in up to 20 exposed kernels per compact. One of the main goals of this test is to assess the retention of fission products in the exposed kernels and the surrounding graphitic matrix material comprising the compact. Future heating tests will expose AGR-3/4 compacts at different temperatures in the FACS furnace (Demkowicz 2017). Additionally, the ability may be developed to heat re-irradiated AGR-3/4 compacts to allow a determination of the retention of short-lived fission products I-131 and Xe-133.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

AGC-2 Specimen Position Adjustment

This engineering calculations and analysis report (ECAR) calculates the Advanced Graphite Creep (AGC)-2 specimen position adjustments that result from (1) as-built dimension corrections of the Advanced Test Reactor (ATR) reactor, experiment, and specimens, (2) thermal growth of the capsule wall and graphite components, and (3) variable specimen elevations as the specimen stacks shrink due to radiation -induced shrinkage and axial load- induced creep. The specimen elevations listed in the experiment assembly documents do not accurately represent the actual specimen positions in relation to the reactor core mid-plane. Prior experiment insertion difficulties have shown that the distance from the reactor top head closure plate to the core is greater than the nominal elevations shown on the ATR facility drawings. Drawing 600001 lists the measured difference between the closure plate and the gearbox support beam as 0.313 in.ches greater than the facility drawings. Thus, the height of the specimens in relation to the core mid-plane is 0.313 in.ches higher than the nominal dimensions. AlsoIn addition, the overall dimension of the experiment is 0.084 in. shorter than the nominal drawing dimension on Drawing 601266 [1]. As a further adjustment, the average weld shrinkage of the three capsule welds is 0.035 each, which affects the as-built position of the lower isolation weld plate (i.e., Part #601266-9) onto which all graphite components are stacked. In addition to the as-built adjustments mentioned above, the specimens also are subject to a stack up error, which is due to their the thickness difference from nominal. This effect is most apparent in the center S-7 stack which that has a large number of specimens (i.e., 170 specimens). Per post-irradiation examination (PIE) measurements, the average S-7 specimen is about 0.0008 in. smaller than nominal. Even though this difference is small, when multiplied by the number specimens, the top specimen experiences an elevation reduction of 0.14 inches. Although dimensional changes were less pronounced in S-1 through S-6, specimen elevations were adjusted based on pre-irradiation length measurements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

AGR-5/6/7 Experiment Preliminary Findings

AGR-5/6/7 in-pile data have been published: Fuel irradiation conditions, Fission gas release-rate-to-birth-rate (R/B) ratios, PIE and safety testing are still in progress and only limited data have been published, and preliminary results.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

AGR-5/6/7 Ceramography Report

The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program was established in 2002 to conduct research and development on tri-structural isotropic (TRISO)-coated particle fuel for High Temperature Gas-cooled Reactors (HTGRs). All AGR irradiations have been completed: AGR-1 (Collin 2015), AGR-2 (Collin 2014), AGR-3/4 (Collin 2016) and AGR-5/6/7 (Pham 2021). The objectives of each program are presented in Figure 1.1. The purpose of the AGR-5/6/7 program was to provide a baseline fuel qualification data set to support licensing, deployment, and operation of HTGRs in the United States. To achieve these goals, the program includes fuel fabrication, irradiations of TRISO fuels and high-temperature materials, safety testing and post-irradiation examination (PIE), fuel performance modeling, and fission product (FP) transport (Sharp 2020). The AGR-5/6/7 work continues as a final irradiation program of the Advanced Reactor Technologies (ART) program, that began in February 2018 and ended in July 2020. The AGR-5/6/7 fuel compacts were irradiated for a total of approximately 360.9 effective full power days (EFPDs) (Stempien 2023), resulting in final burn-up values, on a per-compact basis, ranging from 5.66% to 15.26% fissions per initial heavy metal atom (FIMA), and fast fluence values ranging from 1.62×1025 n/m2 to 5.55×1025 n/m2 (E >0.18 MeV) (Stempien 2022)

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Integrated Approach to Post-Irradiation Examination of Nuclear Materials at Idaho National Laboratory

Idaho national Laboratory (INL) is the U.S. lead national laboratory for the Department of Energy’s Office of Nuclear Energy (DOE-NE), providing much of the nuclear research, development and demonstration capability needed to move nuclear innovation forward to deployment. INL’s Materials and Fuels Complex hosts a unique combination of personnel, facilities and infrastructure and offers the ability to perform post-irradiation examinations (PIE) of nuclear materials spanning multiple length scales. The ability to combine engineering-scale analysis and sub-microscopic characterization provides valuable insights into fuel and structural material behavior and degradation mechanisms. The holistic approach is used to accelerate these materials demonstration and deployment. Selected studies will be presented, highlighting the impact of these techniques on improving fuel reliability, safety, and efficiency, thereby advancing the development of sustainable and advanced nuclear energy technologies.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Why Vascular Surgeons and Interventional Radiologists Collaborate or Compete: A Look at Endovascular Stent Placements

PurposeTo understand how cultural differences between vascular surgeons (VSs) and interventional radiologists (IRs) affect their clinical decision making and inter-specialty relationships.MethodsTwenty-four conversational interviews were conducted with IRs and VSs about their approaches to patient care, views of their specialty and others, and solutions to any expressed concerns. Interview transcripts were systematically analyzed to identify and compare key themes according to the constructivist grounded theory and content analysis using NVivo 10 software. These data were supplemented with a retrospective analysis of 3658 endovascular stent placements performed at a large medical academic center over 11 years. Aggregate counts were divided by provider specialty, and trends were assessed via correlation coefficients.ResultsEndovascular stent placements were relatively equally divided between IR and VS over 11 years with some variability from placements by cardiology. IRs tend to lay claim to treatments as masters of procedures, whereas VSs base their claims on being masters of the treated diseases, leading to collaboration in some practices and bitter competition in others. The level of perceived competition was most associated with specialists’ awareness of and appreciation for specialty-specific values rather than differences in practice structure/reimbursement.ConclusionsUnderstanding cultural differences between IRs and VSs is imperative for fostering better collaboration to grow shared territory rather than competing for the same slice of the pie.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

Estimating the Contribution of the Nickel to Protium Loading of Full-Length Getters

The full-length getter (FLG) is a critical component of the tritium producing burnable absorber rod (TPBAR), designed to capture tritium produced within the lithium aluminate pellets. However, due to the chemical similarity between protium (¹H) and tritium (³H), the FLG readily absorbs protium, reducing its capacity to absorb tritium and increasing the risk of tritium permeation into the reactor coolant. This study evaluates protium produced from neutron irradiation of nickel plating on the FLG through 5?Ni(n,p) reactions, quantifies its contribution to the total protium observed, and informs models of tritium and hydrogen transport within TPBARs. During irradiation, neutron capture by 58Ni results in the formation of 5?Ni, which undergoes neutron bombardment to produce 4He via 5?Ni(n,a) reactions and protium via 5?Ni(n,p) reactions. The measured helium content post-irradiation provides insight into the neutron capture processes within the getter. The 4He measured within the getter may be useful in determining hydrogen produced by the nickel plating on FLG because 59Ni also produces protium in a 59Ni(n, p) reaction. The contribution of 1H from the nickel in FLG contributed less than 1% of the measured H2 gas in PIE, ranging from 0.005% to 0.614%. These findings refine current knowledge of the protium-tritium interplay in TPBARs and support the development of improved transport models for tritium and hydrogen, ultimately aiding in the optimization of TPBAR design and reactor operations.

Arbova, Dana L.↗

A vacuum ultraviolet photoionization study on the thermal decomposition of ammonium perchlorate

Pyrolysis products of ammonium perchlorate (NH 4 ClO 4 ) at 483 K were monitored on line and in situ via single photon photoionization reflectron time-of-flight spectrometry (PI-ReTOF-MS) in the photon energy range of 9.00–17.50 eV. The photoionization efficiency curves (PIE) of the subliming product molecules were collected and allowed for detection of three class of products containing chlorine, nitrogen, and oxygen including atoms and free radicals. The results found suggest a new insight into possible low-temperature decomposition pathways of NH 4 ClO 4 .

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

Comparison of silver, cesium, and strontium release predictions using PARFUME with results from the AGR-1 irradiation experiment

Here, the PARFUME (PARticle FUel ModEl) code was used to predict the release of fission products silver, cesium, and strontium from tristructural isotropic coated fuel particles and compacts during the first irradiation experiment (AGR-1) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-1 experiment used the fuel compact volume average temperature for each of the 620 days of irradiation to calculate the release of silver, cesium, and strontium from a representative particle for a select number of AGR-1 compacts. Post-irradiation examination measurements provided data on release of these fission products from fuel compacts and fuel particles, and retention of silver in the compacts outside of the silicon carbide (SiC) layer. PARFUME-predicted fractional release of silver, cesium, and strontium was determined and compared to the PIE measurements. For silver, comparisons show a trend of over-prediction at low burnup and under-prediction at high burnup. PARFUME has limitations in the modeling of the temporal and spatial distributions of the temperature and burnup across the compacts, which affects the accuracy of its predictions. Nevertheless, the comparisons on silver release lie in the same order of magnitude. Results show an overall over-prediction of the fractional release of cesium by PARFUME. For particles with failed SiC layers, the over-prediction is by a factor of up to 3, corresponding to a potential over-estimation of the diffusivity in uranium oxycarbide (UCO) by a factor of up to 250. For intact particles, whose release is much lower, the over-prediction is by a factor of up to 100, which could be attributed to an over-estimated diffusivity in SiC by about 40% on average. The release of strontium from intact particles is also over-predicted by PARFUME, which also points towards an over-estimated diffusivity of strontium in either SiC or UCO, or possibly both. The measured strontium fractional release from intact particles varied considerably from compact to compact, making it difficult to assess the effective over-estimation of the diffusivities. Moreover, the release of strontium from particles with failed SiC is difficult to observe experimentally due to the release from intact particles, preventing any conclusions to be made on the accuracy or validity of the PARFUME predictions and the modeled diffusivity of strontium in UCO.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗