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At least 235 records · Page 13
All sourcesNASA NTRSDOE OSTI
DOE OSTI2025

The Transient Reactor Test Facility (TREAT)

TREAT operated from 1959-1994, later refurbished & resumed operation in 2017 to support fuel safety testing Zircaloy-clad graphite/fuel blocks comprise core Virtually any power history possible within ~2000 MJ core transient energy capacity From milliseconds to minutes: Pulses, Ramps, LOCA Fuel motion monitoring system “hodoscope” observes fast neutrons emitted from specimens to track fuel relocation in real time Reactor also can be a neutron source to adjacent radiography facility Experiment vehicle does everything else Safety containment, specimen environment, and instrumentation

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Milestone 1.2.11: H 2 Production from Surrogate Non-Native Corrosion Plumes on Aluminum 6061-T6 Fuel Cladding Surrogates

Thick, localized, “non-native” corrosion plumes have been identified on Advanced Test Reactor fuel elements, raising concern on their impact on the radiolytic formation of molecular hydrogen gas (H 2 ) from aluminum-clad spent nuclear fuel (ASNF) under proposed extended (> 50 years) dry storage conditions. Here, we report our findings on H 2 generation from the gamma irradiation (up to 52 MGy) of surrogate “non-native” corrosion plume coupons: ambient-temperature-corroded (~350 days in water) aluminum alloy 6061 (AA6061) coupons in helium gas environments with ~0% added relative humidity. Additionally, we provide a comparison of proposed ASNF drying techniques— vacuum drying only, vacuum drying + 100 °C for 4 hr, and vacuum drying + 220 °C for 4 hr—on the yield of H 2 from these surrogate systems. The presented data indicates that similar amounts of H 2 (~2 × 10–3 µmol J–1) are formed from gamma irradiated AA6061 coupons corroded under different temperature regimes, i.e., ambient/350 days vs. 90 C/30 days. These findings validate current, complimentary modelling predictions based on high-temperature-corrosion irradiation data only. Further, the application of a heat treatment procedure (100 and 220 °C), in conjunction with vacuum drying, accelerated the rate at which a steady-state H 2 yield was attained, in comparison to vacuum only, due to the removal of H 2 precursors in the form of adsorbed waters. Interestingly, within the confidence limits of our measurements, negligible difference in total H 2 yield was found between the two investigated heat treatment procedures.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2023

Auto Stop-Start Fuel Consumption Benefits

With increasingly stringent regulations mandating the improvement of vehicle fuel economy, automotive manufacturers face growing pressure to develop and implement technologies that improve overall system efficiency. One such technology is an automatic (auto) stop-start feature. Auto stop-start reduces idle time and reduces fuel use by temporarily shutting the engine off when the vehicle comes to a stop and automatically re-starting it when the brake is released, or the accelerator is pressed. As mandated by the U.S. Congress, the U.S. Environmental Protection Agency (EPA) is required to keep the public informed about fuel saving practices. This is done, in partnership with the U.S. Department of Energy (DOE), through the fueleconomy.gov website. The “Fuel-Saving Technologies” and “Gas Mileage Tips” sections of the website are focused on helping the public make informed purchasing decisions and encouraging fuel-saving driving habits. Here, in order to provide users with accurate information about the auto stop-start feature, experiments were conducted to determine its fuel economy effect. Four vehicles were tested both with and without the feature enabled under three test cycles: the Federal Test Procedure (FTP) city fuel economy test, the US06 high acceleration aggressive driving schedule that is often identified as the “Supplemental FTP” driving schedule, and the EPA New York City Cycle (NYCC). The results were compared to measure the fuel economy and consumption effects of using the auto stop-start feature. It was found that the fuel economy improvement varied significantly between drive cycles depending on the amount and percentage of idle time during the test. The largest fuel economy improvements were 7.27% and 26.4% for the FTP and NYCC, respectively.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2021

POST-TEST EXAMINATIONS OF A LOCA SAMPLE FROM AN IRRADIATED HIGH-BURNUP PWR M5 FUEL ROD

A LOCA integral test with high burnup PWR M5 fuel rod was conducted in the Irradiated Fuel Examination Laboratory through the complete LOCA sequence: heating the LOCA sample to 300ºC and pressurizing the internal pressure to 8.27 MPa, heating at 5ºC/s from 300 to 1200ºC, holding in steam for 90s at 1200ºC, cooling at 3ºC/s to 800ºC, followed by water quench and rapid cooling to 100ºC. After LOCA testing, examinations, such as the fuel fragmentation analysis, burst and ballooning characterization, axial strain measurement, and microstructural examinations were performed. Metallographic examinations of an as-irradiated high burnup sample adjacent to the LOCA test sample revealed a bonding layer between the fuel and cladding. The posttest LOCA examinations indicates the corrosion layer formed during normal operations in the commercial reactor might provide a protection against the steam oxidation at high temperatures for test times performed in this work. The microstructure of the as-irradiated fuel is compared to the microstructure of the post-LOCA test fuel. Posttest LOCA examination on unirradiated post-tests Zr cladding samples was conducted, which served as baseline data for in cell testing with irradiated samples. The results obtained with irradiated PWR high burnup M5 fuel rad were compared to the LOCA test data obtained with irradiated BWR high burnup Zircaloy-2 fuel rod at Argonne national Laboratory.

Yan, Yong↗
DOE OSTI2020

Fission Gas Monitoring System Efficiency Calibration Summary Report

The Fission Gas Monitoring System was designed to work in conjunction with the Fuel Accident Condition Simulator Furnace and is used to measure fission gas releases during post-irradiation fuel heating tests. A high-purity helium sweep gas is swept past the heated fuel sample in the furnace and routed to a fission gas monitoring system, which cryrogenically traps the krypton (Kr) and xenon (Xe) fission gases. As the radioactive species of the trapped gas decays, high-purity germanium detectors monitor the amount of fission gas in each trap. The primary fission product present in this test is 85Kr, which has a 10.76-year half-life. The presence of 85Kr indicates particle defects and iodine release. There is also the potential for re-irradiating the fuel compacts in the Neutron Radiography Reactor in the Hot Fuel Examination Facility basement. With re-irradiation, it is possible to have xenon fission gas present, specifically 133Xe, which has a 5.243-day half-life. Re-irradiation also produces 131I, which is important for design-basis accidents and regulatory licensing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Summary of Thermocouple Performance in the Advanced Gas Reactor Experiment AGR-5/6/7 During Irradiation in the Advanced Test Reactor

This is a conference presentation ppt file. The following abstract is for the entire conference paper - which will later be the subject of a PRS artifact The Advanced Gas Reactor -5/6/7 (AGR-5/6/7) experiment was the fourth and final experiment in the AGR experiment series and completed irradiation in July 2020. It serves as the formal fuel qualification test for the TRISO fuels under development by the US Department of Energy. This experiment was designed to irradiate fuel particles at temperatures ranging from 800°C ? 1500°C. The high end of the range created unique challenges for thermocouple-based temperature measurements. Commercially available high-temperature platinum-rhodium thermocouples (Types S, R, and B) and tungsten-rhenium thermocouples (Type C) suffer rapid decalibration due to transmutation of the thermoelements from neutron absorption. A special low neutron cross-section thermocouple system based on molybdenum/niobium thermoelements called High Temperature Irradiation Resistant thermocouple (HTIR-TC) has been under development at INL since circa 2004. Several examples of this thermocouple type were incorporated into regions of the test operating above 1200°C. For regions in the test operating in the 1000°C ? 1200°C range, high-performance versions of Type N thermocouples recently developed at Cambridge University were installed. Standard Type N thermocouples were used in regions of the test operating below 1000°C. A total of 54 thermocouples were incorporated into the test. A report of the performance of this large heterogenous set of thermocouples over the first third of the irradiation was previously provided (ANIMMA 2019). This follow-on paper discusses results from the balance of the experiment (Feb 2019 - July 2020). Special attention is focused on potential drift of thermocouples operating in the higher temperature regions.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Determination of Scale Bar for AGHCF Metallography Data

The DOE Nuclear Energy Advanced Reactor Technologies (ART) Fast Reactor Program (FRP) has supported the development of several databases containing information on the safety performance of fast reactors, components, and fuels. This growing collection of legacy experimental data, operating data, and analysis is available online to registered users. Metallography data represents one of the most critical types of post-irradiation examination (PIE) data being collected, organized, and archived in several ART Fast Reactor Databases (https://frdb.ne.anl.gov), including the Metallic Fuels Irradiation & Physics Database (FIPD), Out-of-Pile Transient Database (OPTD), and TREAT (the Transient Reactor Test Facility) Experimental Relational Database (TREXR). These databases contain three principal sets of metallography data. The first set comprises metallography data from Experimental Breeder Reactor-II (EBR-II) and Fast Flux Test Facility (FFTF) irradiated fuel pins examined in the Hot Fuel Examination Facility (HFEF). The second set consists of metallography data from EBR-II irradiated fuel pins examined in the Alpha-Gamma Hot Cell Facility (AGHCF). The third set includes metallography data from transient-tested fuel pins (including both out-of-pile furnace tests and TREAT tests) examined in AGHCF. Since both the second and third sets were generated in AGHCF, they are governed by identical specifications. The metallography data in the databases consist of digital images scanned from either positive or negative photographic films. To analyze the microstructure of a fuel pin, a series of preparatory steps are required, including sectioning, epoxy mounting, mechanical grinding and polishing, and etching. Following sample preparation, specimens are transferred for metallographic examination. The AGHCF and HFEF metallography data were generated using optical microscopes manufactured by Leitz and Bausch and Lomb (B&L). Images were recorded on Polaroid film at preset magnifications. Magnification verification for the Leitz and B&L metallographs was conducted every two months prior to 1989 and at least every six months from 1989 through the conclusion of the IFR program. Magnifications determined from imaging of microslide standards were compared to the instrument settings for magnifications ranging from 50× to 500×. If the magnifications determined from standards deviated from the instrument settings, adjustments were made to the bellows extension until agreement was achieved. The specifications for AGHCF and HFEF legacy metallography data have been established based on available hard-copy and digital records, most of which have been incorporated into the data repositories associated with FIPD, OPTD, and TREXR. Detailed specifications including hard-copy records, digitized records, cutting diagrams and sectioning schemes, high-magnification photographs, photomosaics (composites), information tags, scale bars, and magnification verification procedures can be found in a separate report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Benchmark Specifications for TREAT Tests M5, M6, and M7

Detailed information describing three TREAT tests performed on metallic fuels has been collected and organized for use as a benchmark in evaluating the performance of metallic fuel models and codes. The tests, designated M5, M6, and M7, subjected EBR-II-irradiated fuel pins to a single type of overpower transient (at prototypical conditions with full coolant flow and an exponential power rise on an 8 second period) in flowing sodium loops. Six fuel pins were tested; five were ternary (U-19Pu-10Zr) alloy fuel clad in D9 with burnup ranging from 0.8 to 9.8 at. %, and one was binary alloy fuel (U-10Zr) clad in HT9 with 2.9 at. % burnup. The information gathered from the test records is expected to be useful for pre-transient characterization of the irradiated fuel pins as well as the transient analysis of the metallic fuel when subjected to severe accident conditions. This report presents benchmark specifications for the M5, M6, and M7 TREAT tests, identifies where primary sources of benchmark-related information can be found, and includes background information to help a user of the data understand their applicability and limitations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Engine Operating Conditions, Fuel Property Effects, and Associated Fuel–Wall Interaction Dependencies of Stochastic Preignition

This work for the Coordinating Research Council (CRC) explores dependencies on the opportunity for fuel to impinge on internal engine surfaces (i.e., fuel–wall impingement) as a function of fuel properties and engine operating conditions and correlates these data with measurements of stochastic preignition (SPI) propensity. SPI rates are directly coupled with laser–induced florescence measurements of dye-doped fuel dilution measurements of the engine lubricant, which provides a surrogate for fuel–wall impingement. Literature suggests that SPI may have several dependencies, one being fuel–wall impingement. However, it remains unknown if fuel-wall impingement is a fundamental predictor and source of SPI or is simply a causational factor of SPI. In this study, these relationships on SPI and fuel-wall impingement are explored using 4 fuels at 8 operating conditions per fuel, for 32 total test points. The fuels were directly injected at two different injection timings: an earlier injection timing that initially targets the piston crown and a later injection timing that targets the cylinder liner. At each injection timing, the engine was operated at both 90°C and 70°C coolant and lubricant temperatures, and 185 and 200 kPa absolute intake manifold pressure. This work serves as an exploratory effort to down select conditions and provide initial fuel properties of interest for a secondary study to explore fuel property specific effects on fuel-wall interaction and SPI propensity. Significant findings from this initial operating condition and fuel property exploratory work are: 1. reduced engine operating coolant and lubricant temperatures, along with 2. retarded injection timings were required to increase SPI propensity. Moreover, at these conditions some fuel specific effects were also observed; specifically, increased ethanol content increased measured dye–wall (i.e., fuel–wall) interaction. However, despite increased dye–wall interaction, the increased volatility of the ethanol containing fuels also reduced the estimated fuel retention in the top-ring zone and associated measured SPI propensity. Thus, the findings of this unique approach to explore relationships between fuel-wall impingement and SPI highlight that SPI propensity is more directly proportional to retained fuel, and not simply fuel–wall impingement. Furthermore, fuel retention was found to be directly influenced by complex fuel property and engine operating condition relationships. Either retarded injection timings and/or increased fuel volatility increased fuel wall-impingement, while less volatile fuels and/or reduced coolant temperatures increased fuel retention. Therefore, for a given operating condition, the data highlights that greater volatile fuels exhibit increased fuel wall impingement without increased fuel retention or SPI propensity, while less volatile fuels could exhibit reduced fuel-wall impingement but increased fuel retention and SPI propensity rates.

33 ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2023

Loss of Coolant High Burnup (LOC-HBu) Program

LOC-HBu is a joint experimental program (JEEP) operating within the Nuclear Energy Agency’s (NEA’s) framework for irradiation experiments II (FIDES-II). LOC-HBu is dedicated to the understanding of light water reactor (LWR) fuel performance at high burnup under loss of coolant accident (LOCA) conditions. LOC-HBu aims to support burnup extension needs by addressing identified R&D priorities to achieve an improved understanding of fuel fragmentation, relocation, and dispersal (FFRD) of HBu fuel during LOCA events. The need for such experiments was identified in the United States by the Fuel Performance Testing Technical Expert Group (FPTTEG) under the Electric Power Research Institute (EPRI)’s Collaborative Research on Advanced Fuel Technologies (CRAFT) framework. A combined in-pile and furnace testing program plan was developed by researchers at Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL) which was reviewed and approved by the FPTTEG Combined TREAT-LOC and SATS LOCA Experiment Plan. The first 4 tests identified in this plan are being put forward by the U.S. Department of Energy (U.S. DOE) and INL for inclusion in the FIDES-II second triannual work period. The data produced under this plan will be used to further validate and confirm existing models and inform future R&D and model development. The experimental program was specifically developed to address data gaps and opportunities identified via detailed review of the existing public knowledge base on LOCA FFRD, as well as reviewing specific experimental development activities regarding prototypic LOCA conditions for light-water reactor (LWR) systems. The program will systematically investigate the impacts of prototypic HBu fuel/cladding thermomechanical behaviors under postulated LWR LOCA conditions not yet fully investigated. These conditions correspond with prototypic decay-energy heatup (DEH) and stored-energy heatup (SEH) conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Automotive fuel cell stack and system efficiency and fuel consumption based on vehicle testing on a chassis dynamometer at minus 18 °C to positive 35 °C temperatures

In this work we present an in-depth laboratory technology assessment of a 2016 Toyota Mirai Fuel Cell (FC) vehicle based on chassis dynamometer testing. The 114.6 kW FC stack has a high dynamic response, which makes this powertrain a FC-dominant hybrid electric vehicle. The measured peak efficiency is 66.0% FC stack and 63.7% FC system with an idle hydrogen flow rate of 4.39 g/hr. The high FC system efficiencies at low loads match typical vehicle power spectrums, resulting in a high average vehicle efficiency of 62% compared to 45% and 23% for a hybrid electric vehicle and a conventional vehicle, respectively. An energy breakdown accounts for the FC stack losses, FC system losses, air compressor loads, and heater loads for different drive cycles and different thermal conditions. The cold-start North American city drive cycle (UDDS) energy consumption values are, respectively, 758, 581, 226, and 321 Wh/km at ambient conditions of -18 degrees C, -7 degrees C, -25 degrees C and 35 degrees C with 850 W/m 2 of solar loading. The FC system shutdown and startup processes at temperatures below the freezing point contribute to the increased hydrogen consumption. Additionally, the raw test data files are available for download, thus providing the research community with a public reference data on a modern production automotive FC system.

08 HYDROGEN↗
DOE OSTI2022

Detailed characterization of a PWR fuel rod at high burnup in support of LOCA testing

Experimental investigations of the fuel microstructure and volatile fission products along the pellet of a high burnup specimen (local burnup 76 GWd/tHM) have been conducted to support future transient testing. Detailed microscopy examinations have been carried out at different length scales. Transmission electron microscopy has highlighted a significant amount of damage across the entire radius with the formation of networks of dislocations. The optical and scanning electron microscopy determined the formation of three zones in the pellet with different characteristics. An intermediate region with no grain subdivision, lower porosity than the central zone porosity and high retained fission gas in nanometric bubbles and in the matrix was present between r/r0 ˜ 0.55 and 0.8. The high retention of gas in this region might suggest that the region will be prone to fine fragmentation, in addition to the HBS. An abrupt transition in the structure was observed at mid radius, with a third region developing from the mid radius to the pellet center. In this part of the pellet, metallic and grey phases with size between hundreds of nanometers and a few micrometers have formed at grain boundaries. The majority of fission gas has been released from the grain matrix and the original grains have polygonised, forming sub-grain domains separated by low-angle grain boundaries. No final explanation can be given for the polygonization occurring in the center, but on the basis of the irradiation history and the analysis of all the post irradiation examination (PIE) data, it is postulated that the polygonization within the original grains is an effect of dynamic recovery occurring at high temperature in the fuel center.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A Fast Reactor Irradiation Experiment Design in the ATR

Modern modeling techniques were used to investigate a proposed method for fast neutron irradiations in an existing thermal-spectrum reactor, the Advanced Test Reactor (ATR). This method builds upon pre-existing ideas, where fast flux is increased by surrounding the specimens with fissionable “booster fuel” but diverges from historical approaches by using an already developed fuel element design used in the Belgian Reactor 2 (BR2) as the booster fuel while leveraging modern 3-D modeling and simulation techniques. Design evaluations and neutronics simulations were performed to evaluate the performance of a BR2 fuel element irradiated in an ATR flux trap with test pins in the central channel of the BR2 fuel element. These efforts have yielded promising results. Adding a BR2 fuel element in the northeast (NE) flux trap of ATR was predicted to result in a 150% increase to the incident fast neutron flux with a fast (>0.1 MeV) to thermal (<0.625 eV) neutron flux ratio ranging from approximately 50 to 150, dependent on the material used for thermal neutron filtering and volume of moderator within the central channel of the BR2 element. The predicted annual fast neutron fluence (>0.1 MeV) ranges from 7.9 × 1021 to 9.1 × 1021 n/cm2. Given the relatively large fast to thermal neutron flux ratio, the calculated radial power profiles within 4.3 mm outer diameter U10Zr fueled specimens irradiated within the BR2 booster fuel element are adequate representations of those within fast neutron reactors. The predicted radial power profiles are not flat, but they are more prototypic than those seen in advanced fuels tests which began in ATR in 2003. Another distinct advantage of this experiment design is that full-scale test pins can be irradiated to augment the ongoing series of reduced scale advanced fuels tests. The proposed experiment design irradiated within a BR2 fuel element in a flux trap of ATR offers an improved alternative to the current testing of advanced reactor fuels in ATR. Selected results from this design evaluation are presented.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Neutronics analysis of Full Size Plate–1 Experiment using MC21

Qualification and testing of Low Enriched Uranium fuel is desired so that it can be used in research reactors instead of High Enriched Uranium. The Full Size Plate - 1 (FSP-1) experiment involves irradiating a low enriched metallic fuel in the Advanced Test Reactor to assist in this general qualification. Prior to irradiation, an in-depth neutronics analysis is needed to ensure the safety requirements of the Advanced Test Reactor are met. To accomplish this, the MC21 software was used to analyze the FSP-1 experiment, which will allow a test run to be performed with this new fuel in the Advanced Test Reactor Critical facility. By verifying and validating the results of the MC21 software with comparisons to other codes and the preliminary irradiation in the Advanced Test Reactor Critical facility, the FSP-1 experiment will be ready to irradiate using the Advanced Test Reactor. This work is concerned with the neutronics analysis of the FSP-1 experiment using the MC21 software.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Forced flow transient safety analysis of irradiation device with adjustable orifice for research reactor fuel assemblies

The Belgium Reactor 2 (BR2) of the Belgian Nuclear Research Centre (SCK CEN) has several irradiation devices or rigs that are dedicated to the fuel performance and qualification demonstration testing of research reactor fuels. In support of the U.S. High Performance Research Reactor (USHPRR) LEU conversion project, a new flexible irradiation apparatus, MUSTANG-R, has been constructed. SCK CEN has completed the design and safety study, in cooperation with Idaho National Laboratory (INL) and Argonne National Laboratory (ANL), to allow for the irradiation testing of a full-size fuel assembly in a 200 mm diameter channel in the BR2 reactor. The moveable valve is a key design feature of the device and acts like an adjustable orifice enhancing or restricting the flow through a coolant channel inlet located in the BR2 upper plenum. This moveable valve allows the flow through the device to be adjusted prior to each BR2 cycle to obtain the necessary conditions for the fuel qualification test. This ensures accurate and representative thermal-hydraulic conditions of the fuel design are achieved. The device was designed and qualified as passively safe, implying verification by a combination of mechanical and thermal-hydraulic analysis and testing. This includes characterization of the safety margin required for a scenario where the moveable valve is assumed to be erroneously closed during irradiation. A simplified and conservative method is proposed for analyzing the corresponding forced flow transient using a critical heat flux criterion. In conclusion, this allows the required minimum valve opening to be determined for the experiments' design and safety studies.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

In-Rod Sensor System Overview, Benefits and Recent Irradiation Test Results

Westinghouse is developing an In-Rod Sensor (IRS) System capable of measuring critical fuel rod parameters such as center-line fuel temperature, rod internal pressure and axial fuel pellet stack elongation without having penetrations to the fuel rod. The technology is similar to the Halden Reactor instrumentation, but the IRS system focuses on commercial fuel rod instrumentation. Both systems circumvent fuel rod penetrations, however the IRS system overcomes unique geometric challenges found in commercial fuel rod assemblies. The IRS could also play a key role in accelerated fuel qualification by allowing real-time measurements during separate effects tests, integral fuel testing or during irradiation of lead test rods or lead test assemblies. Combining these measurements with data acquired from targeted separate effects testing and multi-scale modelling could significantly reduce the time required to qualify new fuel and reduce the safety margin uncertainty of operating plants. The IRS is actively in development for Light Water Reactors (LWR) based fuel assemblies but can also be applied to Gen IV reactor designs with minimal modifications.

Carvajal, Jorge↗
DOE OSTI2023

$\mathrm{EBR-II MOX}$ Fuel Characterization Enabling ARES Phase I Testing

Pretransient characterization was performed for the Experimental Breeder Reactor II (EBR-II) mixed-oxide (MOX) fuel pellets from the SPA-2/-2B Operational Reliability Testing collaboration between Japan and the United States. Continued collaboration under the Advanced Reactor Experiments for Sodium Fast Reactor Fuels project will investigate the transient performance of these rods in the Transient Reactor Test facility at Idaho National Laboratory in the MOXTOP-THOR experiment. The results will fill a gap in existing transient performance data for MOX as these rods have a peak burnup of 14.3 at. % (~134.4 GWd/t) in the EBR-II. Fuel pellet properties were gathered from available resources and their irradiation and decay history evaluated. Further reactor physics calculations were performed to support the experiment design, reactor operations, and safety analyses necessary to enable the programmatic success of this effort. Of the three irradiated fuel pins, two will undergo transient testing, and all three will undergo post-irradiation examination. The methodology development and analysis activities utilized we report enable current experiment design work and provide the pathway through which measured data of this type can be further evaluated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗