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Electric Heated Tests were successfully completed; renewing this capability and important measurement type. YH1.9 has now been used in an integral experiment; a monumental step for DOE NE. Preliminary results show feedback in the experiment was typically less than predicted.
A modern cold triple axis spectrometer is being planned for the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory. Here, we describe the design of an incident beamline that will put a flux of ~ 10 8 $\frac {n}{cm^{2}s}$ on a sample with an area of 2 cm X 2 cm. It takes current physical constraints at HFIR into account and it can accommodate both single and multiplexed analyzer-detector secondary spectrometers and large superconducting magnets. The proposed incident beamline includes a multi-channel guide with horizontal focusing, a neutron velocity selector, components to facilitate an incident beam polarization option, and a double-focusing pyrolytic graphite monochromator. Here, this work describes the process of optimizing the guide system and monochromator and summarizes the expected performance of the incident beamline for non-polarized operation.
The Gear Test Assembly (GTA) has completed Campaign #2 of in-sodium testing at the Mechanisms Engineering Test Loop (METL). Campaign #1 was conducted in 2019 and completed 19,600 cycles, correlating to 9,800 fuel assembly maneuvers. Campaign #1 was terminated following a shaft seizure caused by a failed bearing in the sodium gearbox. Campaign #2 was conducted in the spring of 2020 and completed 2,768 cycles, correlating to 1,384 fuel assembly maneuvers. Campaign #2 was terminated following elevated vibration measurements coming from the sodium gearbox. Following cleaning and disassembly, a bearing was found to have failed in the sodium gearbox. Campaign #1 used tapered roller bearings made of 52100 steel with a heat treatment from the manufacturer that allows for operation to 350°C. Campaign #2 used the same tapered roller bearings, but with no heat treatment. A new dynamic shaft seal design was used in Campaign #2 that led to significant reductions in argon consumption and graphite debris generation. A modified external gearbox system led to no external bearing failures and allowed for higher torque output. Finally, modifications to the Carbonation System led to a more effective carbonation process and all-around reduction in turnaround time.
Sufficient AGN-201M fuel plate material exists at Y-12 to support the development of a new, inherently safe, subcritical assembly for use in NCS training courses at ORNL. Four experiments with the ORNL subcritical assembly are possible and feasible. Future work includes the final design of the core and reflector and split table component design specifications. The ORNL Material Demonstration Facility (MDF) may be used to 3D print ORSA stand/table components and graphite reflector. Researchers will consider fuel coatings for contamination control, rather than an Al can. A visit to Y-12 will be made to determine the location of AGN fuel plates and remaining fuel transportation considerations. Final facility location will be considered
Uranium alloyed with 50 wt% zirconium (U-50Zr) is a proposed light water reactor (LWR) nuclear fuel by Lightbridge Corporation (LTBR). The proposed method for making the U-50Zr alloy is to arc-melt master alloys and then remelt in a vacuum induction melter (VIM) to consolidate the material and cast into an intermediate shape. After casting, the material will eventually need to be formed into a desired fuel shape. The work discussed in this report resulted from a joint effort between Ltbr and Pacific Northwest National Laboratory (PNN) to investigate a 500g – 1kg scale casting process to produce the U-50Zr alloy in the desired δ-UZr 2 phase and characterize the impurities and microstructure that result from the casting process. Master alloys were fabricated in an arc melter, then five castings were carried out in a VIM with multiple inert coating and crucible materials to find an appropriate combination. Both ZrO 2 and graphite crucibles were used and different combinations of Y 2 O 3 , CaZrO 3 , and TiC to identify which would contain the molten metal with the least reaction. On each casting, the C, O, N, H impurities were analyzed as well as the phase by x-ray diffraction and microstructure. Of the five castings, two resulted in majority of δ-UZr 2 phase-pure material and had impurity levels within acceptable ranges. The two most successful castings utilized a graphite crucible with a TiC undercoating and a Y 2 O 3 overcoat. The O and N levels were below 1000 ppm and the C content was variable but did not result in measurable carbide formation. The highest success casting resulted in an average of 282 ppm C, 567 ppm O, 217 ppm N and 79 ppm H. This casting's crucible and inert coating material was repeated with slightly different casting parameters and resulted in higher C numbers but similar phase identification. The differences between each casting are discussed and recommendations are made for future experiments to better decide on a casting process to go forward with.
Horizontal Compact High Temperature Gas Reactor (HC-HTGR) is being designed by a multi-disciplinary team of nuclear, mechanical, and structural engineers under the support of a DOE-NE Advanced Reactor Demonstration Program’s Advanced Reactor Concepts-20 (ARC-20) award. The objective of this ARC-20 project is to deliver a conceptual design for the proposed HC-HTGR in 3 years and support its commercialization as a safe, low-cost HTGR. Argonne National Laboratory (Argonne) is responsible for the design and analysis of the reactor cavity cooling system (RCCS) as a safety system for passive decay heat removal of the reactor concept. Additionally, Argonne is providing analysis of the primary coolant system to ensure temperatures within the core remain below safety margins during steady-state and potential accident scenarios. This fourth semi-annual report summarized the progress made at Argonne on the two tasks during the second half of FY23. As a part of the RCCS design task, recent efforts have been made to complete a conceptual design of the RCCS for the HC-HTGR, including the design update of the water panel and system configuration favorable in point of view of fabrication and system operation. Design calculations were conducted under various heat load conditions to validate the system design. Transient simulations using RELAP5-3D were conducted to investigate system dynamics under transients of interest and to evaluate the system performance in the design condition. The results demonstrated the overall system feasibility that the RCCS design maintains structures temperatures lower than maximum allowable temperature with sufficient system inventory without any active heat sink. In the primary system thermal hydraulics task, the preliminary analysis was performed for a long term pressurized conduction cooldown (PCC) transient. This analysis used a combination of a fully resolved and coarse homogenized mesh to predict the temperature distribution for the steady-state initial condition and the PCC transient. The steady-state initial condition was determined using a fully resolved full core model with 3D solid to 1D fluid coupling. The fully resolved mesh was also used to model the first 20 seconds of the PCC. The temperature difference between fuel pins and the graphite matrix becomes minimal and the dominant heat transfer shifts to a larger scale radially towards the RCCS. After 20 seconds, a homogenized coarse mesh is used, greatly reducing the computational costs of the model. These results demonstrate that the core is designed to passively remove enough decay heat in a protected loss of primary coolant flow to prevent an unsafe rise of core temperatures.
Criticality safety analyses are completed on transportation casks used for microreactor whole core shipment to provide examples of models and analyses to industry, regulators, and nuclear community at large to be used in verification and validation analyses of similar applications. The microreactors considered are based on a Gas-Cooled Microreactor (GCMR) and a Heat-Pipe Microreactor (HPMR), both utilize HALEU fuel in the form of TRISO particles and various other design options considered in industry microreactor designs. Variant design options of GCMR and HPMR were also investigated to provide a wider application range for each technology. Criticality safety analyses for the GCMR and HPMR packages were performed using the CSAS6 sequence of SCALE 6.3.2 with the ENDF/B-VII.1-based continuous energy neutron libraries. Different scenarios were investigated, including normal operation and water flooded conditions to represent nominal and hypothetical accident scenarios. Sensitivity and similarity analyses are also performed using the TSUNAMI sequence of SCALE 6.3.2, and the similarity analysis uses all the experiments from the ICSBEP Handbook with High Enriched Uranium (HEU), Intermediate and mixed Enriched Uranium (IEU) and Low Enriched Uranium (LEU) systems. Many ICSBEP experiments were found marginally similar, with similarity index (ck) values greater than 0.8, to the GCMR and HPMR cask models, especially in flooded conditions that may be more constraining due to reduced reactivity margins. Some experiments using TRISO fuel particles and graphite moderator, including the recent THETA and Deimos experiments from LANL, together with some IRPhEP experiments, were also used for the similarity analyses. These TRISO-fueled and graphite moderated experiments behaved similarly to the GCMR models, with maximum ck value greater than 0.9, but they are less similar to the HPMR models. This observation indicates that additional critical experiments might be needed to further validate the criticality safety models, especially for heat-pipe based microreactor transport packages and for non-flooded transportation configurations.
This report summarizes modeling of Big-BUSTER free field characterization (FFC) experiments at the Idaho National Laboratory (INL) Transient Reactor Test Facility (TREAT) reactor performed from December 20th, 2023, to February 22nd, 2024. TREAT is an air-cooled graphite moderated research reactor that can be used to study material response to neutron irradiation. The BUSTER (Broad Use Specimen Transient Experiment Rig) refers to a containment module for placing experiments in the reactor core. The Big-BUSTER is an analogous platform for a modified core configuration with a larger experimental cavity.
Graphite anisotropic microstructure Graphite crystal structure and basal planes Response to irradiation and temperature C-axis and a-axis behavior The effect of porosity Nano-and micro-length scale Microscopic to macroscopic response Material property changes Defect length-scale factors Internal stress build-up Irradiation induced dimensional change Temperature effects Stress increases & property changes
The HTR-PROTEUS IRPhEP Handbook benchmarks represent some of the highest quality benchmarks available for systems with TRISO-HALEU fuel, graphite pebbles, graphite reflector, and high neutron leakage. Nevertheless, significant variability in computed eigenvalue results was encountered in HTR-PROTEUS configurations that are relevant for transportation configurations. The team will investigate the source of these variabilities in the original benchmark. In addition, the team will investigate the HTR-PROTEUS subcritical measurements, kinetics data, and potential criticality effects from the introduction of hydrogen in the system (water ingress). The project will generate a benchmark evaluation of additional HTR-PROTEUS measurements that can significantly increase the value of these criticality benchmarks in testing nuclear codes and data to support transportation validation needs for industry and the U.S. NRC.
The second Advanced Graphite Creep (AGC-2) experiment was designed to irradiate various types of graphite specimens at a temperature of 600°C. The specimens were irradiated in an instrumented leadout capsule experiment in the south flux trap of the ATR during cycles 149A, 1498, 1508, 151A, and 1518. Temperature was monitored using twelve thermocouples located at various elevations in the reactor core, and a helium-argon gas mixture was used for gas gap temperature control of the specimens. The purpose of this analysis is to calculate specimen temperature using measured data on reactor power and helium-argon gas flows, and as-run calculations of heating rates and displacement per atom (DPA) in graphite. The accuracy of the model is assessed by comparing measured and calculated thermocouple temperatures. Uncertainty in gas gaps may preclude an accurate temperature calculation. In these cases, adjustments are made to the thermal model in order to reconcile the measured and calculated thermocouple temperature and to ensure the accuracy of the calculated specimen temperature.
The third Advanced Graphite Creep (AGC-3) experiment was designed to irradiate various types of graphite specimens at a temperature of 900°C. The specimens were irradiated in an instrumented leadout capsule experiment in the east flux trap of the ATR during cycles 152B, 154B, 155A, and 155B. Temperature was monitored using twelve thermocouples located at various elevations in the reactor core, and a helium-argon gas mixture was used for gas gap temperature control of the specimens. The purpose of this analysis is to calculate specimen temperature using measured data on reactor power and helium-argon gas flows, and as-run calculations of heating rates and displacement per atom (DPA) in graphite. The accuracy of the model is assessed by comparing measured and calculated thermocouple temperatures. Uncertainty in gas gaps may preclude an accurate temperature calculation. In these cases, adjustments are made to the thermal model in order to reconcile the measured and calculated thermocouple temperature and to ensure the accuracy of the calculated specimen temperature.
This presentation covers motivations and DNCSH, individual particle measurements, infinite lattice study, application study, and conclusions.
We report the formation mechanism and compositions of a solid-electrolyte interphase (SEI) on a microporous carbon/sulfur (MC/S) cathode in Li–S batteries using a carbonate-based electrolyte (1 M LiPF 6 in ethylene carbonate (EC)/dimethyl carbonate, v:v = 1:1). Through characterizations using 1D and 2D solution-phase nuclear magnetic resonance spectroscopy, coupled with model chemical reactions and DFT calculations, we have identified two critical roles of Li + in steering the SEI formation. First, the preferential solvation of Li + by EC in the mixed carbonate electrolyte renders EC as the dominant participant in the SEI formation, and second, Li + coordination to the EC carbonyl alters activation barriers and changes the reaction pathways relative to Na + . The main organic components in the SEI are identified as lithium ethylene monocarbonate and lithium methyl carbonate, which are virtually identical to those formed on Li and graphite anodes of lithium-ion batteries but via a different pathway.
Carbide-based fuel assembly includes outer structural member of ceramic matrix composite material (e.g., SiC—SiC composite), insulation layer of porous refractory ceramic material (e.g., zirconium carbide with open-cell foam structure or fibrous zirconium carbide), and interior structural member of refractory ceramic-graphite composite material (e.g., zirconium carbide-graphite or niobium carbide-graphite). Spacer structures between various layers provide a defined and controlled spacing relationship. A fuel element bundle positioned between support meshes includes a plurality of distributively arranged fuel elements or a solid, unitary fuel element with coolant channels, each having a fuel composition including high assay, low enriched uranium (HALEU). Fuel assemblies are distributively arranged in a moderator block and the upper end of the outer structural member is attached to a metallic inlet tube for hydrogen propellant and the lower end of the outer structural member is interfaced with a support plate, forming a nuclear thermal propulsion reactor.
This report is a major update of the report ”Cubical moderators for the Neutron Target Demonstator (NTD), LA-UR-23-33899”. It serves as a documentation of milestone M1 of the LDRD project 20240004DR.
A series of fuel irradiation experiments have been planned in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to support the licensing and operation of the Advanced Reactor Technologies high temperature gas-cooled reactor. The advanced gas reactor (AGR) experiments are comprised of multiple independent capsules containing multiple cylindrical fuel compacts, placed inside of a graphite cylinder in ATR. The purpose of the AGR experiments is to provide data on fuel performance under irradiation, support fuel process development, qualify the fuel for normal operating conditions, provide irradiated fuel for accident testing, and support the development of fuel performance and fission product transport models. The advanced graphite creep (AGC) experiments provide irradiation creep data for design and licensing. To date, six irradiation campaigns have been completed: AGR-1 (December, 2006 – November, 2009); AGR-2 (June, 2010 – October, 2013); AGR-3/4 (December, 2011 – April, 2014); AGC-1 (September, 2009 – January, 2011); AGC-2 (April, 2011 – May, 2012); and AGC-3 (November, 2012 – April, 2014).
This Engineering Calculations Analysis Report (ECAR) documents the results of the Advanced Test Reactor (ATR) detailed physics analyses performed to calculate the displacements per atom (DPA) and the fast neutron fluence (E > 0.1 MeV) of the Advanced Graphite Creep (AGC) experiment, AGC-3, irradiated in the ATR East Flux Trap (EFT) (see Figure 1) during ATR Cycle 152B, Cycle 154B, Cycle 155A, and Cycle 155B. This ECAR also reports the neutron and photon heat rates for the materials of the AGC-3 experiment for ATR Cycle 152B. The results for these evaluations and analysis are reported herein.