TREAT Accomplishments FY21
No abstract. These highlight slides will be included on the website https://transient.inl.gov/SitePages/Home.aspx as information of work accomplished.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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Sirius Irradiation Experiments and Post-Irradiation Examinations for Nuclear Thermal Propulsion
Nuclear Thermal Propulsion (NTP) systems hold promise in reducing transit times for exploration of the solar system by both crewed and uncrewed missions. NTP systems currently under investigation include a once-through high temperature gas-cooled fission reactor to provide thermal energy to heat the coolant which also serves as the propellant. The fuel systems of angular UN fuel particles dispersed in a matrix of W/Re, creating a ceramic and metallic composite or cermet, has been irradiated in Idaho National Laboratory’s (INL) Transient Reactor Test Facility (TREAT) enabling evaluation of these materials under representative nuclear heating rates (~95 K/s) and peak temperatures (~2527 K). These tests named Sirius-1 (UN-W/Re), have been irradiated and this paper will present post-irradiation examination results. The Sirius-1 test produced cracks in the fuel specimen and spalling of surface material. Uranium soot was found on the inner wall of the irradiation capsule indicating loss of some fissile material from the fuel specimen. Spalling from the surfaces was also noted upon visual inspection. Uranium diffusion from the fuel particles resulted in the formation of U/Re phases and edge features producing a laminar microstructure.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Cracking in first ramp to power of standard versus high thermal conductivity UO2 pellets via internal nuclear heating
Advanced fuel designs that incorporate thinner fuel UO2 pellets interspaced by high thermal conductivity inserts have been proposed, with the primary goals of reducing peak centerline temperatures and temperature gradients across fuel pellets and enhancing heat transfer from the fuel to the coolant. An initial series of experiments has been performed on this design, including laboratory experiments and a series of experiments using the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) Facility, the latter of which compared thermal gradient driven fracture of standard pellet designs with that in the proposed advanced fuel design. Although reducing fracture is not the primary objective of the new fuel design, the lower thermal gradients are expected to reduce fracture, so it can serve as an indicator of the thermal behavior of this fuel in the reactor. The in-reactor tests were conducted at multiple linear heat generation rates and confirm the expected result that fracture in both the standard and advanced fuel pellets occurs during the first ramp to power in standard light-water reactor conditions. Post-irradiation examination of the experiment material was performed and included quantification of the extent of fracture in the fuel pellets. It was found that the advanced-design pellets reduce the extent of fracture in a statistically significant way. This confirms the expected behavior predicted by two-dimensional axisymmetric models of this experiment. This study is an important first experimental confirmation of the efficacy of the proposed inserts for achieving their desired effect on the thermal behavior of the fuel.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Performance of the HERA Pre-Hydrided Experiments
Presentation on the status of the HERA Pre-hydrided Experiments performed at TREAT to be presented at the Advanced Fuels Campaign Annual Program Review Meeting.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Fuel Performance Simulation of High Burnup Fuels in Planned Integral Design Basis Accident Experiments
High burnup (HBu) fuel rods from the Byron Nuclear Generating Station (BNGS) were recently received at Idaho National Laboratory (INL) to support a variety of planned Nuclear Energy fuel cycle R&D objectives ranging from fuel performance, fuel recycle, and spent fuel research topics. Among these R&D activities, these fuel rods will be the subjects of multiple in-pile experiment programs at the Transient Reactor Test (TREAT) facility as well as detailed characterization and testing in the hot cells at INL and Oak Ridge National Laboratory (ORNL). TREAT RIA experiments are planned for the Nuclear Energy Agency Framework for Irradiation Experiments (FIDES) Joint Experimental Program called High burnup Experiments in Reactivity Initiated Accident (HERA) program. TREAT and ORNL-furnace LOCA experiments are part of the Department of Energy (DOE) Advanced Fuels Campaign (AFC) program U.S. consensus LOCA test plan, and the in-pile experiments have also been proposed in a FIDES project called Loss of Coolant-High Burnup (LOC-HBu). The results of these test programs will provide crucial data about safety performance enabling extended licensable burnup limits for these fuels. The purpose of this paper is to document fuel performance computational simulations of the BNGS fuel using the Bison code. The detailed assessments include (1) the irradiation history of the fuel to provide prediction of as-run fuel conditions and (2) extending the irradiated fuel conditions into the as-designed experiment conditions for the HERA-HBu RIA experiments and for the LOC-HBu LOCA experiments. The results of these assessments will inform post-irradiation examinations (PIE) of the BNGS parent rods and detailed final design of the planned experiments.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
The Automatic Reactor Control System (ARCS) Upgrade, version 2.0 for TREAT
The Transient REActor Test Facility (TREAT) recently underwent an upgraded to the Automatic Reactor Control System (ARCS). The purpose of the upgrade was to institute a new software architecture that is better suited for the programming environment, patched software bugs and applied two new segments for reactor control. This paper will focus on the new segments for reactor control. The two new control segments added to ARCS are called Generic Power and Generic Rods. The original version of ARCS provided only two power related functions, namely periods and ramps, that had to be spliced together to generate any power shape. The Generic Power segment allows the user to input data points to define the function and then the algorithm does its best to create the desired shape. The original ARCS program utilized two direct rod commands (i.e. open loop control), these were a rod stop and clip. A Generic Rod segment extends the ability of the computer to directly command any rod position. The functionality of these new segments was proven during the TREAT outage in 2024.
46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
Preliminary analysis of TREAT free-field experiments using OpenMC
This work analyses activation calculations for dosimetry materials during a steady-state irradiation in the Transient Reactor Test (TREAT) reactor core. Hence, we developed a workflow based on the Monte Carlo code OpenMC alongside a custom depletion solver. The irradiation-induced activity as a function of time is computed, and several sensitivity studies are performed to evaluate uncertainty. This study has shown activity computations are sensitive to flux amplitude, irradiation time, atoms quantity and microscopic cross sections. Stochastic uncertainties have been propagated to evaluate the activity uncertainty for each dosimetry material. Most uncertainties are below our target of 3%, which demonstrates OpenMC as a powerful predictive and analysis tool. The precise results obtained through this newly developed computation scheme will be used in future experiments to characterize quantities of interest when operating the TREAT reactor in new configurations.
61 - RADIATION PROTECTION AND DOSIMETRY↗
ECAR-6440 Rev 2 MARVEL Hazard Evaluation
This hazard evaluation supports the Microreactor Applications Research, Validation, and Evaluation (MARVEL) Project’s design effort and the development of the preliminary documented safety analysis (PDSA) (addendum to safety analysis report [SAR]-420, “Transient Reactor Test [TREAT] Facility FSAR”) (SAR-420-ADD-1). The hazard evaluation process for the MARVEL project for compliance with the requirements in 10 CFR 830, “Nuclear Safety Management,” follows a process similar to the Licensing Modernization Project (LMP) as outlined in Nuclear Energy Institute (NEI)-18-04, “Risk-Informed Performance-Based Guidance for Non-Light Water Reactor Licensing Basis Development,” and supporting documents. The LMP process is adapted to fit Department of Energy (DOE) reactor regulatory requirements as applicable and appropriate using a graded approach based on the MARVEL microreactor design. This approach provides reasonable assurance of meeting the requirements of 10 CFR 830 for protection of the public, workers, and environment.
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
SPC-70525 Rev 2 Beryllium Metal Procurement Specification
Idaho National Laboratory (INL) is developing a microreactor to produce electrical power utilizing a small nuclear core under the Microreactor Application Research Validation and Evaluation (MARVEL) program. Key to this design are neutron reflector components constructed of Beryllium Metal (S-65) which is a structural grade of hot-pressed beryllium block. MARVEL will be located at the Transient Reactor Test (TREAT) facility. Any conflict between this specification and referenced Codes and Standards, or any supplementary specifications in the procurement documents requires written clarification from the Contractor prior to proceeding with any work. Any deviation from the procurement documents requires approval by the Contractor with the change request process.
21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
Transient Liquefaction on the Lunar or Martian Surface Operational Demonstration
As developments of local production of cryogenic fluids on the Lunar or Martian Surface progress, it is important to understand transient system responses to help with the balancing of process plant power and understanding system level operations. During the Cryogenic Fluid In-situ Liquefaction for Landers (CryoFILL) testing, a series of transient oxygen liquefaction tests were completed. These tests included varying liquefaction flow rate, environmental temperature, tank fill level, and effective cryocooler lift while allowing the tank pressure to respond to the input controls. An additional transient test was run at the 90% fill level to determine the impact of injecting the gaseous oxygen at the bottom of the tank, allowing the vapor to bubble up through the liquid. Tests were run in a cyclical nature varying one variable at a time. The control variable was set in a manner to increase tank pressure for a period of time and then subsequently changed in a manner to decrease the tank pressure back to its original value with multiple cycles run for all tests. Tank pressure and system temperature responses were tracked as a function of time with an emphasis on repeatability. Results indicate that of the four variables tested, the environmental temperature is the least important. As expected, the bubbling of the liquefaction gas significantly decreased the pressurization and depressurization rates in the tank at the 90% full level.
Liquefaction↗
Development of NASA's Sample Cartridge Assembly: Summary of GEDS Design, Development Testing, and Thermal Analyses
Outline: Background of ISS (International Space Station) Material Science Research Rack; NASA SCA (Sample Cartridge Assembly) Design; GEDS (Gravitational Effects in Distortion in Sintering) Experiment Ampoule Design; Development Testing Summary; Thermal Modeling and Analysis. Summary: GEDS design development challenging (GEDS Ampoule design developed through MUGS (Microgravity) testing; Short duration transient sample processing; Unable to measure sample temperatures); MUGS Development testing used to gather data (Actual LGF (Low Gradient Furnace)-like furnace response; Provided sample for sintering evaluation); Transient thermal model integral to successful GEDS experiment (Development testing provided furnace response; PI (Performance Indicator) evaluation of sintering anchored model evaluation of processing durations; Thermal transient model used to determine flight SCA sample processing profiles).
O'Connor, Brian↗
Testing Instrument Extremes in the TREAT Facility
TREAT is a transient power-shaping reactor whose primary mission is to enhance safety performance by testing nuclear fuels and materials under thermodynamic and neutronic conditions ranging from off-normal to extreme. Transient testing of nuclear fuels is analogous to car crash testing, in that the dramatic changes commonly seen when comparing the test specimens' initial and final state points necessitate in-situ measuring in order to interpret and understand the evolution of the experiment. TREAT’s experiment design strategy utilizes a highly reconfigurable and accessible reactor core that affords flexibility when installing experiment devices, thanks to the lack of a primary coolant boundary and the inclusion of multiple access points in the reactor bioshield. Within the experiment devices, the test specimens (which are integrated with their test environments), desired instrumentation diagnostics, and safety considerations are all contained in a single engineered package. The experiments encompass various types of environments (e.g., water, sodium, and gas at temperature and pressure), as well as a range of control options (e.g., static or flowing coolants). The experimental conditions are uniquely representative of advanced reactors, providing challenges to sensor performance but also affording a unique opportunity for development and qualification of sensors suited to these environments.
46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
Ring Model Development and Validation for Prismatic HTGR Core Thermal-hydraulics and Safety Analysis
Because of the complex core geometry, prismatic high temperature gas-cooled reactors (prismatic HTGRs) often exhibit complex thermal fluid behaviors during both normal operating and transient conditions. Most HTGR designs rely on passive safety system for decay heat removal, such as the reactor cavity cooling system (RCCS). During postulated accidents like Pressurized Conduction Cooldown (PCC) event, the decay heat is first radially transferred from the core region to the reactor vessel outer surface, then to the RCCS cooling panels. The peak fuel temperature is controlled by heat transfer mechanisms with two distinctive characteristic length scales, i.e., the core-wise effective heat conduction and local heat conduction in the fuel pellet scale. As both length scales are essential to determine the fuel temperature, from the modeling perspective, computer codes must be able to capture heat transfer in both scales. This is challenging for both computational fluid dynamics (CFD) tools because of extremely large amount of computation resources required, and for system analysis codes because of the challenge to model the complex core geometry. Under the support of DOE-NE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, efforts have been pursued to support HTGR technology development and its modeling and simulation needs. There is a particular need for advanced modeling and simulation tools to predict thermal-fluid behavior during safety-related transients. In our previous studies, the ring model was adopted in SAM and further developed to simulate the normal operating condition and a PCC event using the MHTGR-350 design of General Atomics as the reference design. This current work represents a continuation of these previous efforts, and the focus is to critically review and examine simplifications and assumptions made to develop the ring model, and to perform code validation using experimental data from an integral-effect test facility, the High Temperature Test Facility (HTTF) at the Oregon State University. In this study, the test PG-27 from the HTTF test suite was selected for code validation purpose. The test PG-27 is a transient test designed to simulate the PCC event of the MHTGR design. Very good agreements between SAM prediction and experimental measurements were found in both coolant and solid structure temperatures during the transient.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Radiation-Induced Transient Effects in Near Infrared Focal Plane Arrays
This viewgraph presentation describes a test simulate the transient effects of cosmic ray impacts on near infrared focal plane arrays. The objectives of the test are to: 1) Characterize proton single events as function of energy and angle of incidence; 2) Measure charge spread (crosstalk) to adjacent pixels; 3) Assess transient recovery time.
Reed, Robert A.↗
Determination of the effects of heating on modal characteristics of an aluminum plate with application to hypersonic vehicles
The structural integrity of proposed high speed aircraft can be seriously affected by the extremely high surface temperatures and large temperature gradients throughout the vehicle's structure. Variations in the structure's elastic characteristics as a result of thermal effects can be seen by changes in vibration characteristics. Analysis codes that predict these changes must be correlated and verified with experimental data. Analytical and experimental modal test results are given from uniform, nonuniform, and transient thermoelastic vibration tests of a 12 x 50 x 0.19 aluminum plate. The data show the effect of heat on the modal characteristics of the plate. The results showed that frequencies decreased, damping increased, and mode shapes remained unchanged as the temperature of the plate increased. Analytical predictions provided good correlation with experimental results.
Snyder, H. Todd↗
Tantalum capacitor behavior under fast transient overvoltages
Tantalum capacitors were tested to determine failure time when subjected to short-duration, high-voltage surges caused by lightning strikes. Lightning is of concern to NASA because of possible damage to critical spacecraft circuits. The test was designed to determine the minimum time for tantalum capacitor failure and the amount of overvoltage a capacitor could survive, without permanent damage, in 100 microseconds. All tested exhibited good recovery from the transient one-shot pulses with no failure at any voltage, forward or reverse, in less than 25 microseconds.
Zill, J. A.↗
Phasor-Domain Modeling of Inverter Based Resources Considering Unbalanced Grid Faults
A significant increase in inverter-based distributed energy resources (DERs) is expected in distribution systems of the near future. The latest phasor modeling approach of DERs with emerging control technologies can capture relevant dynamics in large-scale three-phase distribution systems with unbalanced construction. However, there is a lack of appropriate phasor models to deal with unbalanced fault scenarios. As such, this paper describes per-phase phasor modeling approach appropriate for unbalanced grid disturbance studies in distribution systems with high penetration of DERs. The overall modeling performance of representative DERs is simulated in phasor domain GridLAB-D environment and validated against controller hardware-in-the loop (CHIL) testing or electromagnetic transient (EMT) based switching model simulation in PSCAD with test cases including different unbalanced grid faults. The high accuracy retained from the simulation verifies the effectiveness of the proposed phasor modeling approach considering interactions between positive and negative sequence signals.
Chen, Zhe↗
Modeling and Analysis of Structural Dynamics for a One-Tenth Scale Model NGST Sunshield
New modeling and analysis techniques have been developed for predicting the dynamic behavior of the Next Generation Space Telescope (NGST) sunshield. The sunshield consists of multiple layers of pretensioned, thin-film membranes supported by deployable booms. Modeling the structural dynamic behavior of the sunshield is a challenging aspect of the problem due to the effects of membrane wrinkling. A finite element model of the sunshield was developed using an approximate engineering approach, the cable network method, to account for membrane wrinkling effects. Ground testing of a one-tenth scale model of the NGST sunshield were carried out to provide data for validating the analytical model. A series of analyses were performed to predict the behavior of the sunshield under the ground test conditions. Modal analyses were performed to predict the frequencies and mode shapes of the test article and transient response analyses were completed to simulate impulse excitation tests. Comparison was made between analytical predictions and test measurements for the dynamic behavior of the sunshield. In general, the results show good agreement with the analytical model correctly predicting the approximate frequency and mode shapes for the significant structural modes.
Johnston, John↗