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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Pressure Drop in Seven-Pin Wire-Wrapped Rod Bundle for the Sodium Cartridge Loop in Versatile Test Reactor
This work studies the hydrodynamics of the seven pin wire-wrapped rod bundle in the sodium cartridge loop for the Versatile Test Reactor (VTR) through scaled water experiments and computational fluid dynamics (CFD) simulations. Here, the scaling analysis is first performed to demonstrate the hydrodynamic similarity between water and sodium flows at the same Reynolds number Re. A separate-effects test facility is designed and constructed based on the scaling analysis. Detailed experimental data on the pressure drop covering a wide range of Re values (1165 to 27 689) are obtained, which are used to evaluate existing correlations for friction factor and to benchmark CFD simulations. The experimentally determined friction factors f Exp agree well with the Upgraded Cheng and Todreas Detailed Correlation and Pacio-Chen-Todreas Detailed Model within ±7% but are significantly underpredicted by Rehme's correlation by 25%. Various CFD near-wall treatment methods are tested using ANSYS Fluent and evaluated by experimental data. It is found that when the recommended wall y + values are met, most of the near-wall treatment methods can give accurate friction factor predictions. The resolved near-wall method (y + ~ 1) with the Shear Stress Transport k - ω turbulence model and the scalable wall functions (y + > 12) with the realizable k - ε turbulence model can predict f Exp within ±5% The standard wall functions (y + > 30) and nonequilibrium wall functions (y + > 30) with the realizable k - ε model can predict f Exp within ±10%.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Carburization and decarburization behavior of Grade 91 ferritic-martensitic steel in liquid sodium environments
Here, this paper presents a study of carbon transfer and its effect on microstructure and tensile properties of Grade 91 (G91) ferritic-martensitic steel exposed to sodium at 550–650 °C. Sodium exposure tests were conducted in Argonne's forced convection sodium loops up to exposure times of ~40,000 h. Thermal aging study of G91 steel was conducted in parallel to isolate the thermal aging effect from the sodium effect. It was found that sodium exposures at 650 °C dissolved M 23 C 6 carbides, eliminated the martensite subgrain structure resulting in excessive grain growth and reduced the tensile strength by >50%, while sodium exposures at 550 and 600 °C had an insignificant effect on its microstructure and tensile properties. These effects were attributed to the carburization/decarburization process of G91 steel in sodium environments. Carbon concentrations in sodium were determined by a foil equilibration method. The estimated carbon concentration was in the range of 0.8–1.2 ppm in the SMT-1 loop and 0.3–0.7 ppm in the SMT-2 loop. Thermodynamic analysis of the carburization – decarburization process was conducted for G91 steels exposed in sodium environments. The carbon activity-concentration relationship for G91 was evaluated by considering four phases in G91, i.e. bcc ferrite, M 23 C 6 , NbC and VC carbides. It was found that the carburization-decarburization process in G91 steel was dictated by M 23 C 6 carbides at high carbon activities, while NbC and VC carbides dominated the process at low carbon activities. The calculated carburization-decarburization boundary showed that G91 would undergo decarburization at 650 °C and carburization at 550 °C in the sodium loop environments, which was consistent with our experimental observations. This experimental and theoretical analysis provided a basis for predicting the effect of carbon transfer on the integrity of reactor components in sodium environments and for the design of new alloys used in sodium-cooled fast reactors.
36 MATERIALS SCIENCE↗
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↗
Thermal hydraulic characteristics of liquid sodium and supercritical CO 2 in a diffusion bonded heat exchanger
Compact heat exchangers are being examined for use in advanced, high-temperature nuclear or concentrated solar power plants to improve thermal efficiencies and reduce capital costs. At the time of this work, no experimental investigation of liquid sodium in the popular diffusion-bonded heat exchanger design had been performed. Herein this paper describes the thermal-hydraulic and overall operations performance observed during testing of a 316/L stainless steel “zig-zag” microchannel heat exchanger operating between a (0.5 kg/s) sodium loop and a (0.3 kg/s) recuperated supercritical CO 2 system with outlet conditions up to 550 °C at 16 MPa. Observed pressure drop is presented in the form of an effective friction factor correlation which aligns closely with recent experimental data. The thermal performance of this heat exchanger, measured by its overall conductance under various flow conditions, was decomposed to examine the independent convection behaviors of both the CO 2 and sodium streams. These results are presented in the form of Nusselt correlations fitted to the experimental data. The sodium heat transfer correlation aligns closely with the widely accepted Lubarsky-Kaufman correlation, while the CO 2 heat transfer correlation lies between previous correlations developed by computational and experimental methods. This was the first long-duration use of sodium in a microchannels (~1.8 mm hydraulic diameter) heat exchanger. The use of an oxide control system enabled trouble-free operation of the heat exchanger for over 150 h of high-temperature operation. This first-of-a-kind demonstration using liquid sodium and supercritical carbon dioxide shows great promise for further application of this technology in advanced nuclear or solar power plants.
42 ENGINEERING↗
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↗
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.↗
Development of a Sodium Fast Reactor Cartridge Loop Testing Capability for the Versatile Test Reactor
Not Available
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Progress Report of Alloy 709 Performance in Liquid Sodium
With improved high-temperature properties, Alloy 709 shows great potential for advanced reactor applications. To understand the effects of sodium on A709, small tensile specimens were machined from two commercial heats and subject to sodium exposure in two forced convection loops operated by Argonne. Thermal aging experiments were also conducted in parallel to isolate any sodium-specific effects. An increasing trend in yield strength and a decrease trend in ultimate tensile strength can be seen with the increasing exposure time, suggesting microstructural changes at these temperatures. Nonetheless, the thermally aged and sodium exposed samples behaved similarly, implying no sodium-specific effect at these temperatures. In contrast, 316H SS exposed to sodium at 650°C showed a strong sodium effect. The different responses of A709 and 316H SS to sodium exposure is of interest and requires additional effort to understand. The cold traps of the sodium loop were replaced early in FY23. During a maintenance operation, an unexpected sodium leak incident occurred at one of the sodium loops, triggering an emergency response. An investigation into the incident revealed that the leak happened at a VCR fitting on the recently replaced cold trap. The metal gasket of the leaked VCR fitting was not compressed sufficiently. A dimensional mismatch between the new component and old piping configuration was responsible for a significant misalignment. A new cold trap with improved design is in production. The new design provides additional flexibility and an access point for post-installation helium leak check and for venting of trapped gas. These design features will make the system more error-tolerant and the installation quality verifiable. A thermodynamic analysis was performed to understand the carburization-decarburization behavior of A709 in sodium environment. The equilibrium carbon activity in the alloy was evaluated with M 23 C 6 , TiC, and NbC. The carburization-decarburization boundary was calculated with different methods developed previously. The results suggests that A709 will be carburized in SMT-1 and SMT-2, and can be decarburized in low-carbon-activity environments such as EBR-II at temperatures above 620-650°C. Nonetheless, A709 should be less likely to be decarburized than 316H SS when exposed to the same low-carbon-activity environment.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Chemical Looping of Manganese to Synthesize Ammonia at Atmospheric Pressure: Sodium as Promoter
Abstract Affordable synthetic ammonia (NH 3 ) enables the production of nearly half of the food we eat and is emerging as a renewable energy carrier. Sodium‐promoted chemical looping NH 3 synthesis at atmospheric pressure using manganese (Mn) is here demonstrated. The looping process may be advantageous when inexpensive renewable hydrogen from electrolysis is available. Avoiding the high pressure of the Haber‐Bosch process by chemical looping using earth‐abundant materials may reduce capital cost, facilitate intermittent operation, and allow operation in geographic areas where infrastructure is less sophisticated. At this early stage, the data suggest that 0.28 m 3 of a 50 % porosity solid Mn bed may suffice to produce 100 kg NH 3 per day by chemical looping, with abundant opportunities for improvement.
Aframehr, Wrya Mohammadi↗
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↗
MARVEL Guard Vessel System FEA and ASME Analysis
The INL is seeking to build the first microreactor in the United States, and the first advanced reactor in over 50 years. With the Microreactor Applications Research, Validation & EvaLuation (MARVEL) Project, this microreactor should be operational within a significantly faster timescale than has ever been seen in the nuclear industry. It uses two passive-convection loops, primarily Sodium Potassium eutectic (NaK), and Ga-In-Sn eutectic as an intermediate fluid. Due to the high reactivity of NaK, the primary coolant loop requires a secondary confinement boundary to prevent uncovering of the core in the event of a NaK leak. This secondary boundary is mostly accomplished through the addition of the Guard Vessel, which is a bucket-like metal containment, which is welded onto the primary boundary to maintain an inert gas environment and pressure
99 GENERAL AND MISCELLANEOUS↗
Single-phase, natural circulation annular flow measurements for cartridge loop irradiation experiments
The nuclear industry is increasingly considering cartridge-style experiments for irradiation testing of advanced reactor fuels and materials under flowing conditions. Cartridge loops do not require the extensive support infrastructure that are necessary for external flow loops and minimize the possibility of coolant solidification over the long distance from the reactor to the external facilities. However, there is a general lack of quality flow data for internally heated fluids in an annular configuration representative of a cartridge-type irradiation experiment, particularly one with natural circulation. To address this data need, a series of experiments was conducted to measure the natural circulation flow rates of pressurized water in a sealed, internally heated vessel with annular flow conditions that represent a molten salt or sodium cartridge loop. Temperatures and flow rates were measured under steady-state and transient conditions. Here, this paper describes the facility, methods, and results of the experiments, including the determination of nondimensional parameters. A simple 1D model of the natural convection flow rates agrees well with the experimental results. Applying this model to simulate a liquid salt cartridge experiment predicts that natural circulation flow might be able to provide liquid salt Reynolds numbers similar to those of some molten salt reactor concepts at relevant power densities.
42 ENGINEERING↗
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↗