Cb-1zr Pumped Sodium Loop
Cb-1Zr pumped sodium loop tests to evaluate loop components, EM pump, pressure transducers, and instrumentation techniques
Hoffman, E. E.↗
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HotCell Components
To support operations in the HFEF Hot Cell and the FCF Hot Cell, multiple components were designed to support the Bottle Cutter Project and the Sodium Loop Project. These components were designed and analyzed to ensure they can function in their environment and be operated remotely. FEA analysis were performed, prototypes were made, and parts were manufactured to complete this project. This poster documents these parts and test results.
11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Sodium Chemistry Assessment and Initiation of Sodium Exposure Experiment on A709 Commercial Heats and LPBF 316H SS
This report documents sodium chemistry characterization and the initiation of forced‑convection sodium exposure experiments for two structural materials for sodium‑cooled fast reactors (SFRs): Alloy 709 (solution‑annealed and with an additional 775 °C/10 h precipitation treatment) and laser powder bed fusion (LPBF) 316H stainless steel tensile specimens printed with an as‑built surface condition. The sodium exposure experiment was conducted in a forced‑convection sodium loop with two vessels operated at 550 °C and 600 °C, and a cold trap maintained at <130 °C. Carbon activity in sodium was quantified using a nickel‑foil equilibration method. Initial post‑exposure evaluation after ~1,000 h indicates negligible thickness change for Alloy 709 and a small but measurable mass decrease. To isolate thermal effects from sodium effects, parallel thermal‑aging experiments were performed on Alloy 709 at 550 °C and 600 °C. Baseline tensile tests at room temperature, 550 °C, and 600 °C show expected reduction in strength with increasing temperature and serration tensile curves at elevated temperatures, consistent with dynamic strain aging. After 3,000 h thermal aging, Alloy 709 exhibits modest strengthening and reduced ductility at 550 °C. Together, these results establish the loop chemistry conditions and provide the initial corrosion and mechanical property results for ongoing long‑duration sodium exposure of Alloy 709 and LPBF 316H stainless steel.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Cb-1Zr sodium thermal convection loop
Design, fabrication and testing of Cb-1Zr thermal convection loop for evaluation of refractory metals in boiling and condensing potassium environments
Hoffman, E. E.↗
SNAP-8 simulator loop mechanical design
SNAP 8 simulator using primary sodium-potassium loop, mercury loop operating on Rankine cycle, and heat rejection sodium-potassium loop
Reznik, J. F.↗
SNAP-8 third loop optimization
Eutectic sodium potassium and OS-124 considered as coolant fluids for SNAP-8 third loop - optimum loop operating parameters
EUTECTIC ALLOY↗
A Conceptual Design Study on the Application of Liquid Metal Heat Transfer Technology to the Solar Thermal Power Plant
Alkali metal heat transfer technology was used in the development of conceptual designs for the transport and storage of sensible and latent heat thermal energy in distributed concentrator, solar Stirling power conversion systems at a power level of 15 kWe per unit. Both liquid metal pumped loop and heat pipe thermal transport were considered; system configurations included: (1) an integrated, focal mounted sodium heat pipe solar receiver (HPSR) with latent heat thermal energy storage; (2) a liquid sodium pumped loop with the latent heat storage, Stirling engine-generator, pump and valves located on the back side of the concentrator; and (3) similar pumped loops serving several concentrators with more centralized power conversion and storage. The focus mounted HPSR was most efficient, lightest and lowest in estimated cost. Design confirmation testing indicated satisfactory performance at all angles of inclination of the primary heat pipes to be used in the solar receiver.
Zimmerman, W. F.↗
Potassium corrosion test loop development - Purification, analysis and handling of sodium and potassium Topical report no. 4
Corrosion test loop, and purification, analysis, and handling of sodium and potassium
SODIUM↗
Laser-spectroscopy testbed for impurity monitoring in liquid metal-cooled fast reactors
A significant challenge in sodium-cooled fast reactors is controlling impurities, in particular oxygen impurities, within the sodium coolant, as they can accelerate corrosion and indicate leaks. Optical methods offer the potential to rapidly detect small concentrations of both gaseous and metal impurities that accelerate corrosion, plug coolant channels, and lead to increased activation of isotopes in the coolant. We present the design and performance of an apparatus designed to enable the application of multiple optical analytical techniques, such as laser-induced breakdown spectroscopy, to detect elemental impurities in the sodium melt with high sensitivity. Here, we experimentally demonstrate the detection of characteristic sodium and oxygen spectral lines in liquid sodium, which sets the stage for the optimization of its analytical sensitivity. A robust sensor of this type integrated with the sodium cooling loop has the potential to significantly improve the safety and operational efficiency of generation IV nuclear reactors.
47 OTHER INSTRUMENTATION↗
Performance and evaluation of two liquid-metal pumps for sodium-potassium service
Performance tests on liquid metal pumps for sodium potassium loops
Gorland, S. H.↗
Impurity and liquid level measurements for the Faret loop
Self-seeking liquid sodium level and sodium impurity measurements for fast reactor test loop
SODIUM↗
Auxiliary loop for removal of decay heat from a space power reactor
Heat transfer to auxiliary sodium coolant radiator loop of space power reactor
Petrik, E. J.↗
Benchmark Exercise Report for Experimental Study of Bubble Scrubbing in Water Coolant Pool
Mechanistic assessments of radionuclide release during postulated accidents are expected to be included in advanced reactor license applications. The mechanistic source term (MST) provides an opportunity for vendors to realistically evaluate the radiological consequences of an incident, and may aid in justifying reduced emergency planning zones and plant sites. However, the development of MSTs for advanced nuclear reactors is challenging because there are numerous phenomena that can affect the transport and retention of radionuclides. As part of a trial MST assessment for a metal-fueled, pool-type sodium cooled fast reactor (SFR), led by Argonne National Laboratory, a simplified radionuclide transport code (SRT code) was developed, which includes models to estimate the quantity of fission product aerosols scrubbed in the sodium pool during postulated accident scenarios. In a pool-type SFR, when fission products are released into the coolant pool due to failure of fuel pins, most of the radionuclides are scrubbed by the coolant pool, but some have the potential to migrate to the cover gas region through entrainment within gas bubbles. The SRT code contains a model that evaluates this scrubbing behavior and calculates the fraction of fission product aerosols that reach the cover gas. Due to a lack of available validation data for sodium pool scrubbing, the U.S. Department of Energy funded an experiment at the University of Wisconsin-Madison to measure aerosol scrubbing by injecting air bubbles containing aerosol into a coolant pool. Prior to performing an experiment with liquid sodium, a water loop experiment was performed. Their experiment evaluated the effect of changing the aerosol size, aerosol density, aerosol concentration, bubble size, and pool depth on the aerosol scrubbing efficiency of the pool. In this benchmark experiment, the base tests were conducted by repeated tests of isolated bubbles. Afterwards, more prototypic tests with bubble swarms were performed to evaluate the interactions between the bubbles. The bubble swarm test was able to confirm that a larger amount of aerosol scrubbing occurred than the single bubble test. It was also confirmed that as the bubble size, aerosol density, and pool height increase, the extent of pool scrubbing also increases and does not change with the aerosol concentration. In addition, since the degree of scrubbing is the lowest at aerosol sizes between 0.01 and 1 μm, that is, the largest amount of aerosol is emitted, it was confirmed that the analysis of this size in MST is the most important. This benchmark experiment informs the direction of future sodium experiments.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Work function determination of promising electrode materials for thermionic energy converters
The work function determinations of candidate materials for low temperature (1400 K) thermionics through vacuum emission tests are discussed. Two systems, a vacuum emission test vehicle and a thermionic emission microscope are used for emission measurements. Some nickel and cobalt based super alloys were preliminarily examined. High temperature physical properties and corrosion behavior of some super alloy candidates are presented. The corrosion behavior of sodium is of particular interest since topping cycles might use sodium heat transfer loops. A Marchuk tube was designed for plasma discharge studies with the carbide and possibly some super alloy samples. A series of metal carbides and other alloys were fabricated and tested in a special high temperature mass spectrometer. This information coupled with work function determinations was evaluated in an attempt to learn how electron bonding occurs in transition alloys.
Jacobson, D.↗
Refining NEAMS MARVEL Reactor Model Accuracy
This report outlines the progress of Idaho National Laboratory (INL) in improving the fidelity of the multiphysics model of the Microreactor Applications Research Validation and Evaluation (MARVEL) reactor developed in FY24. The multiphysics model firstly developed in FY24 leverages three single-physics models coupled via the MOOSE’s MultiApps and Transfer systems. The first single-physics model, that functions as the driver application, leverages Griffin to model the neutron transport in the core through the discontinuous finite element (DFEM) discrete ordinates solver (SN). The second single-physics model uses BISON to compute the solid temperature in the reactor. Finally, the System Analysis Module (SAM) was used to model the flow of the sodium-potassium eutectic in the primary loop. In this report we detail the following improvement and enhancements: • An updated mesh leveraging the latest feature in the reactor module (see Section 3.1.).
42 - ENGINEERING↗
Thermal Hydraulic Experimental Test Article: Second Year of Testing with Secondary Sodium System (Fiscal Year 2025 Final Report)
The Thermal Hydraulic Experimental Test Article (THETA) is currently installed in the Mechanisms Engineering Test Loop (METL) 28” test vessel #4. Both the primary and secondary sodium systems remain online to facilitate continued testing. This fiscal year, work was performed using a COMSOL Multiphysics magnetohydrodynamic model to characterize flow more accurately in the secondary electromagnetic flowmeters. Experimental campaigns were then performed to study the thermal hydraulic differences between sodium and water as a surrogate fluid in the THETA geometry as well as a study to better characterize and understand temperature oscillations that exist at the outlet of the core to the hot pool. A peer-reviewed article was published in the ASME Journal of Nuclear Engineering and Radiation Science detailing the THETA facility and providing an overview of a test that was performed with the primary and secondary system online [1]. Work continues to develop a database to house experimental THETA data to better facilitate collaboration with industry and laboratory partners for their use of the data for code benchmarking/validation. THETA remains fully operational and is positioned for continued testing in fiscal year 2026.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Design and Analysis of a Mutual Inductance Level Sensor for Liquid Metals
Here, this article describes the design and analysis of an electromagnetic level sensor for use in high-temperature liquid metal systems. The mutual inductance level sensor (MILS) described in this work was fabricated using two single-conductor mineral insulated cables wrapped in a bifilar fashion around a stainless steel tube core and was housed in an isolating thimble that preserved the pressure boundary of the test vessel. Two sensor variations were fabricated that differ only in active length, 1016 and 1778 mm. Experimental data were collected using the 1016-mm sensor (MILS-MKII-040) that demonstrated a sensitivity of 9.2 μ V/mm in a room temperature testing stand that used solid aluminum as a surrogate for liquid metal. Experimental data were collected using the 1778-mm sensor (MILS-MKII-070) that demonstrated a sensitivity of 6.9 μ V/mm in the high-temperature (300 ° C) liquid sodium environment at the mechanisms engineering test loop (METL) of Argonne National Laboratory. The sensor performance was found to be repeatable over the course of several months, with roughly ±1% deviation from nominal output. Finite element models were developed in COMSOL Multiphysics that fully describe each test setup, and the models were validated using experimental data. The validated COMSOL models were used to perform an array of analyses that examined the performance of the sensor in differing environments. Maximizing the coil diameter inside the isolating thimble was found to maximize the signal and sensitivity of the sensor. An optimal operating frequency was found near 1000 Hz using both experimental data and COMSOL. The influence of a metallic thimble surrounding the sensor and a metallic sensor core was quantified and found to be negligible at the optimal operating frequency. The sensitivity of the sensor was quantified when monitoring the level of additional liquid metals. These include lead, lead-bismuth eutectic (LBE), sodium-potassium alloy (NaK), and lithium (in addition to sodium). The sensitivities were quantified using liquid metal properties at 350 ° C and 650 ° C. The geometry of the test stand model, all material properties used in the model, and the results are presented in a manner that allows the reader can replicate the model and perform additional analyses.
COMSOL↗
Biogenic carbon capture at pulp mills via sodium spiking and oxy-fuel calcination
Over 13 million metric tons of biogenic carbon dioxide (CO 2 ) are mineralized yearly in United States (US) pulp mill recovery boilers as molten sodium carbonate. The mineralized CO 2 is released downstream in the rotary lime kiln where it can be captured at a relatively low cost for permanent sequestration. The use of biocarbon and bioenergy at pulp mills enables the possibility for atmospheric carbon removal when biogenic CO2 is captured and sequestered. We demonstrate the feasibility of capturing biogenic CO 2 via sodium spiking coupled with oxy-fuel calcination in the rotary lime kiln at existing kraft pulp mills. Sodium (Na) spiking elevates the Na ion loading in the kraft chemical looping process by replacing chlorine-based bleaching with a highly alkaline bleaching sequence. Chemical pulping processes are modeled and simulated to understand the technical limits of implementing sodium spiking in existing pulp mills. Each 1 % of sodium added to the bleaching operations increases the rate of CO 2 mineralization by 4 %, and the maximum increase in sodium content in the kraft process is 9 %. Without sodium spiking, estimated CO 2 capture costs are $131 and $107/mt CO 2 for air and oxy-fuel combustion, respectively. With the implementation of sodium spiking, the cost of CO 2 captured decreases by 27 % and 31 %, showing costs of $96 and $74/mt CO 2 for air and oxy-fuel combustion, respectively. Oxy-fuel combustion reduces the costs of CO 2 capture compared to air combustion, but merged deployment with sodium spiking seems to be the most cost-effective pathway to integrate carbon capture.
60 APPLIED LIFE SCIENCES↗