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This report investigates the response of the inverter under different terminal voltage and operating conditions. The goal is to understand the control objective of the inverter (e.g., injection of reactive current for voltage dips) based on the inverter’s response. No attempt is made to determine the exact control algorithms implemented in the inverter.
The state-of-the-practice methods for aerodynamic design of wind turbine blads use linearized blade element momentum theory (BEM) to optimize the twist and chord profiles from a pre-selected set of 2D airfoil shapes. In this work, we apply invertible neural network (INN) tools to enable the rapid inverse aerodynamic design of wind turbine blades including component airfoils. The INN is trained on data obtained through the use of robust automated mesh generation and the HAMSTR computational fluid dynamics solver with advanced turbulence and transition models validated for turbine applications. Our design technique is a significant improvement over the state-of-the-practice linearized blade element momentum (BEM) techniques in capturing 3D nonlinear aerodynamic effects that are critical for optimal design of the rotors. This is made possible by developing sparse, invertible neural networks (INNs) for inverse design and optimization that realize a 100x cost reduction compared to adjoint-based computational fluid dynamics (CFD) approaches, while enabling increased robustness of the final design. We demonstrate the INN tool for design of a section of the NREL 5-MW blade. All generated shapes satisfy the desired aerodynamic characteristics, demonstrating the success of the INN approach for inverse design of wind turbine blades.
Replacing combustion appliances with heat pumps can be part of a strategy to electrify buildings. When their electricity is supplied by renewables, the adoption of heat pumps for space heating may help curb the carbon dioxide emissions inherent to traditional heating systems that burn natural gas or coal. When operating in very cold climates, however, heat pump performance deteriorates, sometimes requiring backup heating supplied by electric resistances or other sources to continue to function. To avoid the use of electric resistances, complementary technologies, like thermal energy storage, solar collectors, or alternative refrigerants, have been explored in recent years. In this study, we present an open-source Python-based numerical model developed to evaluate the annual hourly performance of an air source heat pump (ASHP) operating in the USA. Simulation results include coefficient of performance and heating capacity. The model, validated with experimental results, will be employed in the future to investigate the feasibility of integrating thermal energy storage and other technologies into ASHPs operating in cold climates.
A key issue during the initial start-ups of new Small Modular Reactors (SMRs) is the lack of operational data for reactor model validation. To help better understand the accuracy of the reactor analysis codes CASMO5 and SIMULATE5, higher order comparisons to MCNP5 have been performed. These comparisons are for an initial core conceptual design of the NuScale reactor. The data have been evaluated at Hot Zero Power (HZP) conditions. Comparisons of core reactivity, fuel temperature coefficient (FTC), and moderator temperature coefficients (MTC) have been performed. Comparison results show good agreement between CASMO5/SIMULATE5 and MCNP5 for the conceptual initial core design. (authors)
Helical tube heat exchangers provide better heat transfer characteristics, an improved capability to accommodate stresses due to thermal expansions and a more compact design with respect to straight tube heat exchangers. For these advantages they are considered as an option for the Steam Generator (SG) of many new reactor projects of Generation III+ and Generation IV. In particular, their compactness fits well with the requirements of Small-medium Modular Reactors (SMRs) of integral design, where all the primary system components are located inside the reactor vessel. In this framework, thermal hydraulics of helical pipes has been studied in recent years by Politecnico di Milano in different experimental campaigns. Experiments have been carried out in a full-scale open loop test facility installed at SIET labs in Piacenza (Italy)), to simulate the SG of a typical SMR. The facility includes two helical pipes (1 m coil diameter, 32 m length, 8 m height), connected via lower and upper headers. Following recently completed experimental campaigns dedicated to pressure drops and density wave instabilities, this paper deals with a new experimental campaign focused on both pressure drops (single-phase flow and two-phase flow, laminar and turbulent regimes) and flow instabilities. The availability of a large number of experimental data, in particular on two-phase flow, is of fundamental interest for correlation development, model validation and code assessment. Two-phase pressure drops have been measured in adiabatic conditions, ranging from 200 to 600 kg/m{sup 2}s for the mass flux, from 30 to 60 bar for the pressure and from 0.1 to 1.0 for the flow quality. The channel characteristics mass flow rate - pressure drop has been determined experimentally in the range 10-40 bar, varying the mass flow rate at a fixed value of the thermal flux. In addition, single-phase pressure drops have been measured in both laminar and turbulent conditions. Density wave instabilities have been studied at mass flux from 100 to 400 kg/m{sup 2}s and pressure from 10 to 20 bar, to confirm the particular behavior of the stability boundary in helical geometry at low pressure and low mass flow rate. Finally, starting from the unstable regions identified from the experimental channel characteristics, Ledinegg type instabilities have been investigated to drawn stability maps with complete stable and unstable regions in the dimensionless plane N sub-N pch. (authors)
Spent fuel from domestic and foreign research reactors is received and stored at the Savannah River Site's L Area Material Storage (L Basin) Facility. This DOE-owned fuel consists primarily of highly enriched uranium in metal, oxide or silicide form with aluminum cladding. Upon receipt, the fuel is unloaded and transferred to basin storage awaiting final disposition. Disposition alternatives include processing via the site's H Canyon facility for uranium recovery, or packaging and shipment of the spent fuel to a waste repository. A program has been developed to provide a phased approach for dry storage of the L Basin fuel. The initial phase of the dry storage program will demonstrate loading, drying, and storage of fuel in twelve instrumented canisters to assess fuel performance. After closure, the loaded canisters are transferred to pad-mounted concrete overpacks, similar to those used for dry storage of commercial fuel. Unlike commercial spent fuel, however, the DOE fuel has high enrichment, very low to high burnup, and low decay heat. The aluminum cladding presents unique challenges due to the presence of an oxide layer that forms on the cladding surface, and corrosion degradation resulting from prolonged wet storage. The removal of free and bound water is essential to the prevention of fuel corrosion and radiolytic generation of hydrogen. The demonstration will validate models predicting pressure, temperature, gas generation, and corrosion performance, provide an engineering scale demonstration of fuel handling, drying, leak testing, and canister backfill operations, and establish 'road-ready' storage of fuel that is suitable for offsite repository shipment or retrievable for onsite processing. Implementation of the Phase I demonstration can be completed within three years. Phases II and III, leading to the de-inventory of L Basin, would require an additional 750 canisters and 6-12 years to complete. Transfer of the fuel from basin storage to dry storage requires integration with current facility operations, and selection of equipment that will allow safe operation within the constraints of existing facility conditions. Examples of such constraints that are evaluated and addressed by the dry storage program include limited basin depth, varying fuel lengths up to 4 m, (13 ft), fissile loading limits, canister closure design, post-load drying and closure of the canisters, instrument selection and installation, and movement of the canisters to storage casks. The initial pilot phase restricts the fuels to shorter length fuels that can be loaded to the canister directly underwater; subsequent phases will require use of a shielded transfer system. Removal of the canister from the basin, followed by drying, inerting, closure of the canister, and transfer of the canister to the storage cask are completed with remotely operated equipment and appropriate shielding to reduce personnel radiation exposure. (authors)
A combination of transient heat transfer analytical solutions for a finite line source, a series of point sources, and a series of parallel infinite line sources were combined with a quasi-steady-state multi-layered cylindrical solution to simulate the temperature response of a deep geologic radioactive waste repository with multi-layered natural and engineered barriers. This evaluation was performed to provide information to scientists and decision makers to compare candidate geologic media for a repository (crystalline rock [granite], clay, salt, and deep borehole), and to provide input for the future evaluation of the trade-off between pre-emplacement surface storage time, waste package size, and repository footprint. This approach was selected in favor of the finite element solution typically used to analyze the temperature response because it allowed rapid comparison of a large number of alternative disposal options and design configurations. More than 100 combinations of waste form, geologic environment, repository design configuration, and surface storage times were analyzed and compared. The analytical solution approach used to analyze the repository temperature response allowed rapid comparison of a large number of alternative disposal options and design configurations. More than 100 combinations of waste form, geologic environment, repository design configuration, and surface storage times were analyzed and compared. This approach allowed investigation of the sensitivity of the results to combinations of parameters that show that there is much flexibility to be gained in terms of spent fuel management options by varying a few key parameters. This initial analysis used representative design concepts and thermal constraints based on international design concepts, and it also included waste forms representing future fuel cycles with high burnup fuels. Unlike repository designs with large open tunnels and pre-closure ventilation, all of the disposal concepts analyzed in this study used enclosed emplacement modes, where the waste packages were in direct contact with encapsulating engineered or natural materials. The deep borehole repository concept limits the size of the SNF waste packages and may require rod consolidation to fit within the drill casing diameter. A single assembly waste package, assuming rod consolidation, was evaluated in the current analysis. Similar size restrictions apply for the HLW canisters. At this time no thermal constraints have been defined for the deep borehole repository concept. Representative EBS materials and properties were evaluated. However, changes in EBS design concepts and materials can also have significant effects on the maximum waste package surface temperature. Increased thermal conductivity of the buffer layer can be achieved by using an engineered buffer consisting of a mixture of graphite, sand, and bentonite [14]. One of the advantages of the analytical model is that it highlights the sensitivity of the results to the parameters that define the repository layout, including spacing between axial and lateral neighboring waste packages and drifts. It is clear that repository layout adjustments can be made to reduce the calculated peak temperatures. The results also show that significant reductions in required surface storage times can be achieved if higher thermal constraints can be justified Additional studies are planned to evaluate the trade-offs between surface storage times, repository layout parameters, and variations in EBS design concepts. Model validation and uncertainties will also be addressed. It is expected that shorter surface storage times and more optimized repository design configurations may be achieved. (authors)
A research program is proposed to investigate the behavior of salt when subjected to thermal loads like those that would be present in a high-level waste repository. This research would build upon results of decades of previous salt repository program efforts in the US and Germany and the successful licensing and operation of a repository in salt for disposal of defense transuranic waste. The proposal includes a combination of laboratory-scale investigations, numerical simulations conducted to develop validated models that could be used for future repository design and safety case development, and a thermal field test in an underground salt formation with a configuration that replicates a small portion of a conceptual repository design. Laboratory tests are proposed to measure salt and brine properties across and beyond the range of possible repository conditions. Coupled numerical models will seek to describe phenomenology (thermal, mechanical, and hydrological) observed in the laboratory tests. Finally, the field test will investigate many phenomena that have been variously cited as potential issues for disposal of thermally hot waste in salt, including buoyancy effects and migration of pre-existing trapped brine up the thermal gradient (including vapor phase migration). These studies are proposed to be coordinated and managed by the Carlsbad Field Office of DOE, which is also responsible for the operation of the Waste Isolation Pilot Plant (WIPP) within the Office of Environmental Management. The field test portion of the proposed research would be conducted in experimental areas of the WIPP underground, far from disposal operations. It is believed that such tests may be accomplished using the existing infrastructure of the WIPP repository at a lower cost than if such research were conducted at a commercial salt mine at another location. The phased field test is proposed to be performed over almost a decade, including instrumentation development, several years of measurements during heating and then subsequent cooling periods, and the eventual forensic mining back of the test bed to determine the multi-year behavior of the simulated waste/rock environment. Funding possibilities are described, and prospects for near term start-up are discussed. Mining of the access drifts required to create the test area in the WIPP underground began in November 2011. Because this mining uses existing WIPP infrastructure and labor, it is estimated to take about two years to complete the access drifts. WIPP disposal operations and facility maintenance activities will take priority over the SDI field test area mining. Funding of the SDI proposal was still being considered by DOE's Offices of Environmental Management and Nuclear Energy at the time this paper was written, so no specific estimates of the progress in 2012 have been included. (authors)
The High Burnup Spent Fuel Data Project (herein referred to as the 'project'), sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally defining the effects of long-term storage and transportation on high burnup (HBU) pressurized water reactor (PWR) fuel (>45 gigawatt days per metric ton uranium). The goals of the project are to provide confirmatory data for model validation and potential improvement; to provide input to future dry storage cask design; to support license renewals and new licenses for Independent Spent Fuel Storage Installation facilities; and to support transportation licensing for high burnup used fuel. In support of these project goals, an instrumented TN-32B bolted-lid cask at the North Anna reactor site will be loaded with intact, HBU PWR fuel assemblies (project assemblies) having four different kinds of cladding: standard Zircaloy-4, low-tin Zircaloy-4, ZIRLO, and M5. In approximately 10 years, the instrumented cask will be opened for inspection and several fuel rods from the project assemblies will be characterized. To support the project goals, the project pulled 25 'sister rods', 9 from the project assemblies and 16 from similar HBU assemblies, for characterization. The 25 sister rods are all high burnup and cover the range of cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed nondestructive and destructive examination. Examinations are intended to provide baseline data on the initial physical state of the cladding and fuel prior to the loading, drying, and long-term dry storage process. Further examinations are focused on determining the effects of temperatures encountered during and following drying. Similar tests will be performed on rods taken from the project assemblies at the end of the storage period (the 'cask rods') to identify any significant changes in the fuel rods that may have occurred during the dry storage period. Additionally, some of the sister rods will be used for separate effects testing to expand the applicability of the project data to the fleet, and to address some of the data-related gaps associated with extended storage and subsequent transportation of high burnup fuel. A draft test plan is being developed that describes the experimental work to be conducted on the sister rods. This paper summarizes the draft test plan and necessary coordination activities for the multi-year experimental program to supply data relevant to the assessment of the safety of long-term storage followed by transportation of high burnup used fuel. (authors)
Transient testing involves placing nuclear fuel or material into the core of a nuclear reactor designed to operate at high power for a short time, and subjecting the nuclear fuel or material to short bursts of intense, high power radiation. After the experiment is completed, the fuel or material is analyzed to determine the effects of the radiation. The results are then used in fuel or material design and/or qualification. Transient testing is required for essentially all nuclear fuel design and qualification efforts to learn how nuclear fuel will respond during accidents involving transient overpower and/or under cooling events. For example, nuclear fuel may fragment when exposed to higher than normal power. This fragmentation can cause unacceptable performance. Transient testing is also needed to validate performance models for nuclear fuel and materials. These models, when validated, will dramatically shorten the development and qualification life cycle for nuclear fuels, supporting rapid development of low emissions, reliable power generation. In the past, transient testing was primarily done in facilities that no longer exist. The U.S. Department of Energy evaluated how to provide the required transient testing capability and, following completion of the National Environmental Policy Act process, they selected the Transient Reactor Test (TREAT) facility to resume transient testing. TREAT is located at Idaho National Laboratory (INL). The TREAT reactor was used from 1959 to 1994 to conduct more than 2,500 nuclear fuel transient tests, and was placed in standby in 1994. TREAT is capable of transients in excess of 19 GW. The paper describes the progress towards restart. The restart efforts have revealed that the plant fuel and equipment is acceptable for future operations. The safety basis development, procedures, and training are well under way. Experiment capability preparations are in progress. Reactor startup is anticipated no later than 2018. (author)
The High Burnup Spent Fuel Data Project (the 'project'), sponsored by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally evaluating the effects of long-term dry storage and transportation on high burnup (HBU) pressurized water reactor (PWR) fuel (>45 gigawatt days per metric ton uranium). The goals of the project are to provide confirmatory data for analytical model validation and potential improvement; to provide input to future dry storage cask design; to support license renewals and new licenses for Independent Spent Fuel Storage Installation facilities; and to support transportation licensing for high burnup used fuel. In support of these project goals, an instrumented TN-32B bolted-lid cask (the Research Project Cask or RPC) at the North Anna reactor site will be loaded with intact, HBU PWR fuel assemblies (Project Assemblies) having four different kinds of cladding: standard Zircaloy-4, low-tin Zircaloy-4, ZIRLO{sup R}, and M5{sup R}. In approximately ten years, the instrumented cask will be opened for inspection and several fuel rods from the Project Assemblies will be characterized. The project has also removed 25 'sister rods' from assemblies at North Anna: 9 directly from the project assemblies and 16 from similarly constructed and operated HBU assemblies. The 25 sister rods are all high burnup and cover the range of domestically-used cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed non-destructive examinations (NDE) followed by destructive examination (DE). Examination of the sister rods will provide a baseline for comparison with the cask rods and will allow for separate effects testing. Dry storage canister loading is completed under water in a designated area of the plant's spent fuel pool. In preparation for dry storage, the fuel assemblies and canister cavity must be drained and dried. Typically, the highest temperatures experienced by the fuel during dry storage are encountered either during drying or transfer of the canister to the storage pad; therefore, as a part of the sister rod testing these temperatures will be simulated in the lab to produce samples that are representative of the peak cladding temperature the fuel experiences either during drying, transfer, or storage. (authors)
The nature of computational simulations requires the inclusion of an uncertainty analysis, as we have limited knowledge of all physically determined input parameters for a computationally simulated problem. We rely on uncertainty quantification (UQ) to characterize our confidence in the outcomes. Quantifying uncertainty can provide a basis for certifications in high-consequence decisions, such as nuclear reactor design, and is a fundamental component of model validation. We employ a previously developed method of uncertainty quantification, polynomial chaos expansion with stochastic collocation (PCE-SC), applied to deterministic phonon transport simulations. In these simulations, we use the neutron transport code Rattlesnake, which solves the Self-Adjoint Angular Flux (SAAF) formulation of the transport equation with a continuous finite element (CFEM) spatial discretization and discrete ordinates, spherical harmonics angular discretizations. Rattlesnake was developed in the multi-physics object oriented simulation environment (MOOSE) framework. We have previously shown Rattlesnake to be effective in simulating phonon transport. We aim to provide a deterministic phonon transport framework for heterogeneous nuclear fuel with fission product defects to predict thermal conductivity (?). A first principles, physics-based calculation of thermal conductivity must involve factors such as the microstructure of nuclear fuel, which constantly changes during the fission process through the formation of isotopic decay products. Heat transport in oxide nuclear fuels is dominated by phonon transport. Impurities in the bulk material influence the transport of energy at the fundamental level, altering the scattering behavior of phonons and electrons. Conventionally, heat transport follows classical physics based on the heat equation derived from Fourier's law q = -k∇T, where q is heat flux, k is thermal conductivity and ∇T is a temperature gradient. However, Fourier's law is a macroscopic empirical law in which the thermal conductivity? does not have a mechanistic connection to the underpinning heat transport processes. Thermal conductivity of a material depends both on the material's intrinsic ability to transport heat and a variety of resistive effects caused by defects in the material. Thus ab initio prediction of the macroscopic conductivity - the property of interest for safe reactor operation - requires simulating detailed processes of heat transport, and then determining the effective thermal conductivity of the simulated material from the resulting heat flux under the imposed temperature difference. The thermal conductivity of a bulk, homogeneous, dielectric material can be well estimated by the mechanistically derived expression: k{sub bulk} = 1/3 C{sub ν}νgλ (2), where C{sub ν}, νg, and λ are the volumetric specific heat, phonon speed, and phonon mean free path, respectively. In these problems the relevant parameter is the material's acoustic thickness, its characteristic distance L relative to the phonon mean free path λ. In Eq. (2) the largest source of uncertainty is λ. While propagating uncertainty in λ to uncertainty in k through Eq. (2) is trivial, systems with extrinsic scattering require more sophisticated approaches. At the microscopic level, uncertainty in λ changes both a material's intrinsic thermal conductivity. Our goal is to propagate the uncertainty in λ through the deterministic transport computation of k. Because there are a small number of uncertain input quantities, we use the method of polynomial chaos expansion (PCE), which expresses solutions in the form of spectral expansions of an uncertain variable. This approach combines both intrusive and nonintrusive methods of uncertainty propagation techniques and results in a unique formulation which is very effective and efficient for problems with few uncertain parameters. Uncertainty Quantification and Sensitivity Analysis We use PCE-SC to measure propagation of uncertainty in a 3-D phonon transport problem of homogeneous silicon, to compute the mean and variance in temperature, heat flux and thermal conductivity. (authors)
Geothermal district energy systems (DES), and specifically geothermal networks, provide a viable solution for decarbonizing residential and commercial heating and cooling loads. District energy systems of all kinds enable a thermal resource with a relatively high capital cost (such as a geothermal borehole field) to be shared among a large number of users. While district heating and cooling has been studied for many years, geothermal networks, fifth generation DES that utilize water-source heat pumps and an ambient temperature loop to meet heating and cooling loads, have not been implemented extensively and thus require additional technical and economic optimization to obtain maximum benefits. This paper presents a newly developed reduced-order model that captures the flow of energy within the network, including the commercial and residential users' electrical usage, at an hourly rate over a year. The model includes building loads, heat pumps, borehole fields, and auxiliary heat/cool input, all connected with an ambient-temperature thermal loop model. In the model, operational control is possible for the borehole fields, circulation pump, and auxiliary system. For a given system, the model can output the complete state parameters for each component, the thermal loop, and the collective system, such as flow rate over time, average thermal loop temperature over time, and total electricity usage. The model can also be used to optimize the system control for maximizing system efficiency or minimizing system operational cost. For example, one initial assessment of the borehole controller for an example system showed that a controller with on/off operation of the borehole field reduces annual electrical usage by 33%, compared with continuous operation mode. Hence, the model can assist in optimizing a given system's operation to get the most value out of a geothermal network installation. Future work will consider the model's application to a demonstration project, including the model validation against operational data and system operation optimization.
Geological thermal energy storage (GeoTES) utilizes the underground reservoirs to storage and dispatch energy per given demand schedule over a time scale up to seasons of a year. The energy input can be of various sources/forms; in this paper, we investigate 1) the GeoTES technology with solar thermal hybridization and using depleted oil/gas reservoirs; 2) the GeoTES technology with heat pumps charged by excess renewable electricity and using low-temperature shallow reservoirs. For each GeoTES technology, we carry out the suitability analysis of candidate reservoirs, develop initial techno-economic models, and validate the model with selected case study. The paper provides an overview of our technical progress on the topics of concern and aims to promote a wider acceptance of the GeoTES technologies in the future energy market.
This study presents steam condensation scaled experiments and semi-empirical models in presence of nitrogen (N)—a noncondensable gas (NCG), simulating air in the reactor containment—to support water-cooled small modular reactors (SMRs) passive containment cooling system (PCCS) design and analysis. Previous experimental studies on PCCS are focused on fixed and smaller tube (mostly 2-in.) geometries and specific test condition variations, bringing challenges with geometric scaling and mismatching with SMR prototypic design. To address these challenges, this study presents steam condensation test dataset obtained from three scaled test sections of 1-, 2-, and 4-in.-diameter steam condensers with an annular/jacket cooling of 2-, 3-, and 6 in.-diameter tubes, respectively. Test data were collected for steam ranges from 58 to 63 kg/hr., and NCG flow of 4.4 to 13.3 kg/hr. Annular cooling water flow was varied to obtain required testing conditions of saturated steam inlet and fully condensed outlet. Axial temperature test data of bulk cooling water, steam and condensate were collected by thermocouples for three test sections and various steam-NCG mixing/testing conditions. A standard data reduction method was adopted—utilizing iterative and nodalized mass and heat transfer calculation—to estimate axial local heat fluxes, heat transfer coefficients (HTCs), condensation rates, film thickness, and Nusselt number. Based on the obtained dataset semi-empirical model results—a ratio of experimental and Nusselt’s theoretical HTC are presented. Such results and findings are supportive of developing scaled-up testing facility, to enable model validations and accelerate next generation of reactors development and deployment
Orientation and completion for well pairs that have been subjected to multi-zonal stimulation play a critical role in the long-term performance of an Enhanced Geothermal Reservoir. Here we present the development of a methodology to rapidly and efficiently numerically simulate mixed fracture-matrix flow systems for evaluation of well design and completion options. An example evaluation based on a small fracture network representative of FORGE Well 16(A and B)-78(32) follows a discussion of the theory and model validation.
Pronghorn is a MOOSE based thermal-hydraulics code developed at Idaho National Laboratory (INL) for advanced nuclear reactor analysis. It has been previously applied to model pebble-bed high temperature reactors (HTRs), liquid-metal cooled reactors, and molten salt reactors, among others. This work leverages the coarse-mesh modeling capabilities in Pronghorn to model the Oregon State University (OSU)'s High Temperature Test Facility (HTTF). The HTTF is a 1:4 height scaled-down facility of General Atomics' Modular High Temperature Gas-cooled Reactor (MHTGR). The facility is primarily built to generate data for code and model validation, and does not precisely replicate MHTGR conditions. Nevertheless, it encompasses the main physics associated with MHTGR transients.
This study presents steam condensation scaled experiments and semi-empirical models in presence of nitrogen--a noncondensable gas (NCG), simulating air in the reactor containment--to support water-cooled small modular reactors (SMRs) passive containment cooling system (PCCS) design and analysis. Previous experimental studies on PCCS are focused on fixed and smaller tube (mostly 2-in.) geometries and specific test condition variations, bringing challenges with geometric scaling and mismatching with SMR prototypic design. To address these challenges, this study presents steam condensation test dataset obtained from three scaled test sections of 1-, 2-, and 4-in.-diameter steam condensers with an annular/jacket cooling of 2-, 3-, and 6 in.-diameter tubes, respectively. Test data were collected for steam ranges from 58 to 63 kg/hr., and NCG flow of 4.4 to 13.3 kg/hr. Annular cooling water flow was varied to obtain required testing conditions of saturated steam inlet and fully condensed outlet. Axial temperature test data of bulk cooling water, steam and condensate were collected by thermocouples for three test sections and various steam-NCG mixing/testing conditions. A standard data reduction method was adopted--utilizing iterative and nodalized mass and heat transfer calculation to estimate axial local heat fluxes, heat transfer coefficients (HTCs), condensation rates, film thickness, and Nusselt number. Based on the obtained dataset semi-empirical model results--a ratio of experimental and Nusselt's theoretical HTC are presented. Such results and findings are supportive of developing scaled-up testing facility, to enable model validations and accelerate next generation of reactors development and deployment.