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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Time Analysis on a Steam Condensation Jet

The time dependence of a steam bubble entering a water pool was measured for this experiment. The temperature of the water tank at the nozzle exit was 75 deg. C ± 5 deg. C. The experiment will focus on how the bubble changes its size through time for future experimental analysis concerning the heat transfer coefficient. The study focuses at different steam flow velocities approaching the water tank. The area of the steam bubble was measured using a high-speed camera to obtain the heat transfer coefficients. The region of interest is illuminated with a laser or a halogen lamp depending on the experiment. Calculating the rate at which the bubbles are created and condensed, will help obtain a better understanding of the energy transfer between the bubble and the water. The study of steam condensation is important due to its application on nuclear plants, heat exchangers, geothermal power plants and steam ejectors. Recently, more experiments have been performed and studied due to its high heat transfer rate. The results obtained in this experiment will be used to validate CFD models. This primary experimental report focuses on the steam bubble's time dependence; the same system will be used for future experiments concerning the heat transfer coefficient. In previous work done by Chun using only steam, the average heat transfer coefficient of direct contact condensation reported was in the range of 1.0-3.5 MW/m{sup 2} deg. C. He proposes an empirical correlation to predict the average heat transfer coefficients. The experimental data agrees with the empirical correlation within ± 30%. On a different experiment performed by Gulawani using a hot film anemometer on a steam condensation jet. The reported heat transfer coefficient for the vapor liquid interface varies between 0.1-4 MW/m{sup 2}-K. The nozzle diameter used during this experiment varied between 1-2 mm, and the velocity ranged between 286-304 m/s. Aya and Nariai reported three different heat transfer coefficients for direct contact condensation. They reported a heat transfer coefficient at chugging of 2*10{sup 6} W/m{sup 2}-K, at condensation oscillation of 10{sup 5} - 10{sup 6} W/m{sup 2}-K, and at the surface of stratified flow of 3*10{sup 3} - 3*10{sup 4} W/m{sup 2}-K. With steam and air mixture, a heat transfer coefficient relating the bubble average area was reported by Qu. He reported that the heat transfer coefficient decreased with the increase of the bubble diameter. The relation is time dependant and can be used to the transient analysis of the heat transfer. For this experiment, a set of images was obtained of a steam jet. The pictures were analyzed to obtained the time dependence of the steam bubble with time, and the rate of its condensation. A frequency of 20 Hz was calculated when the steam had a flow rate of 2.5 grams per second. The first result obtained for this experiment. In the future, more flow rates will be analyzed to obtain a relationship between the frequency of the bubbles and the flow rate of the steam. Eventually, this information will aid with the calculation of the heat transfer coefficient of the steam jet in a water pool.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Plant-wide simulation model for transient studies on the XE-100 - 18464

This paper gives an overview of the plant-wide simulation model that is used by X-energy LLC for developing their plant design and control philosophy, as well as assessing plant upset conditions. The simulation model was created in Flownex{sup R} thermal-fluid system simulation software, which solves mass, momentum and energy equations for compressible, incompressible and homogeneous two-phase flow. The plant model encompasses all major components, from reactor to condenser, including main control loop elements. The paper pays particular attention to the validation of the two-phase flow framework used for the secondary steam cycle. Two steady-state validation cases are done using operational data from the THTR. Validation of transient operation is done by simulating density wave oscillations, a type of system-level boiling phenomenon. This is regarded as a thorough test of transient capability because of the challenges this phenomenon poses to several aspects of the calculation and to the consistency of the code as a whole. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Steam Oxidation Testing in the Severe Accident Test Station

Since 2011, Oak Ridge National Laboratory (ORNL) has been conducting high temperature steam oxidation testing of candidate alloys for accident tolerant fuel (ATF) cladding. These concepts are designed to enhance safety margins in light water reactors (LWR) during severe accident scenarios. In the US ATF community, the Severe Accident Test Station (SATS) has been evaluating candidate materials (including coatings) since 2012. Compared to the current UO 2 /Zr-based alloy fuel system, alternative cladding materials need to offer slower oxidation kinetics and a smaller enthalpy of oxidation in order to significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models. However, prior modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. In some cases, the results have been difficult to interpret and more fundamental information is needed such as the stability of alumina in flowing steam at 1400°-1500°C. This report summarizes recent work to measure the steam oxidation kinetics of candidate alloys, the evaporation rate of alumina in steam and the development of integral data on FeCrAl compared to conventional Zr-based cladding.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Steam Oxidation Testing in the Severe Accident Test Station

After the March 2011 accident at Fukushima Daiichi, Oak Ridge National Laboratory (ORNL) began conducting high temperature steam oxidation testing of candidate materials for accident tolerant fuel (ATF) cladding in August 2011 [1-11]. The ATF concept is to enhance safety margins in light water reactors (LWR) during severe accident scenarios by identifying materials with 100× slower steam oxidation rates compared to current Zr-based alloys. In 2012, the ORNL laboratory equipment was expanded and made available to the entire ATF community as the Severe Accident Test Station (SATS) [4,12]. Compared to the current UO2/Zr-based alloy fuel system, an ATF alternative would significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident [13-14]. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models [15-17]. However, initial modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. Also, because many accident scenarios include steadily increasing temperatures, the required data are not traditional isothermal exposures but exposures with varying “ramp” rates. In some cases, the steam oxidation behavior has been surprising and difficult to interpret. Thus, more fundamental information continues to be collected. In addition, more work continues to focus on commercially-manufactured tube material. This report summarizes recent work to characterize the behavior of candidate alloys exposed to high temperature steam, evaluate steam oxidation behavior in various ramp scenarios and continue to collect integral data on FeCrAl compared to conventional Zr-based cladding.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Demonstration of Fully Coupled Calculations in Helical Steam Generator: Toward Predictions of Fluid Elastic Instability

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past seven years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In the last two years a significant effort was devoted to 1) develop and validate a fully coupled capability in SHARP for Fluid-Structure calculations (FSI) and 2) validate the use of SHARP for one-way coupled calculations in helical steam generators on legacy data available as well as novel experiments. In this report we discuss an extension of previous one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. It was demonstrated that in the buffeting regime one-way coupling was judged sufficient because the pressure loads do not cause substantial displacements. However, higher speeds led to a reduction in accuracy and the need for a fully coupled capability for velocity or approaching the critical velocity and fluid elastic instability. In this report we demonstrate such fully coupled capability on the Argonne test case discussed in previous reports.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Steam Condensation Scaled Experiment in the Presence of Non-condensable Gas for Reactor Containment Passive Safety Analysis

This study presents scaled experiments using steam condensation with non-condensable gas (NCG)—helium, simulating hydrogen—as these experiments are pivotal for water-cooled reactor passive containment cooling system (PCCS) design and analysis. Research into PCCSs for small modular reactors (SMRs) is especially important in light of SMR system design; however, studies in the literature reflect limitations due to test geometry and operational condition variations, without considering SMR prototypic design. To address these challenges, a scaled test facility was developed to accurately replicate SMR PCCSs. This facility includes vertical down-flow condensing test sections with 1-, 2-, and 4-in.-diameter condensing tubes, accompanied by annular water cooling. Experiments were conducted using both superheated and saturated steam, with steam mass flow rates varying from 55 to 66 kg/hr., in the presence of helium as the NCG mass flow rate ranges from 1.8 to 22 kg/hr. Test data were collected on (a) the axial temperatures of the annular cooling water; (b) the outer wall temperature of the condensers; and (c) the mass flow rate, temperature, and pressure at the test section inlets and outlets. These primary test data were used in conjunction with a standard data reduction methodology to estimate essential thermal parameters such as heat fluxes, heat transfer coefficients, and condensation rates. The effects of NCGs on steam condensation within the geometry of the scaled test sections were then presented in regard to various testing conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Presentation: Steam Condensation Scaled Experiment in the Presence of Non-condensable Gas for Reactor Containment Passive Safety Analysis

This study presents scaled experiments using steam condensation with non-condensable gas (NCG)—helium, simulating hydrogen—as these experiments are pivotal for water-cooled reactor passive containment cooling system (PCCS) design and analysis. Research into PCCSs for small modular reactors (SMRs) is especially important in light of SMR system design; however, studies in the literature reflect limitations due to test geometry and operational condition variations, without considering SMR prototypic design. To address these challenges, a scaled test facility was developed to accurately replicate SMR PCCSs. This facility includes vertical down-flow condensing test sections with 1-, 2-, and 4-in.-diameter condensing tubes, accompanied by annular water cooling. Experiments were conducted using both superheated and saturated steam, with steam mass flow rates varying from 55 to 66 kg/hr., in the presence of helium as the NCG mass flow rate ranges from 1.8 to 22 kg/hr. Test data were collected on (a) the axial temperatures of the annular cooling water; (b) the outer wall temperature of the condensers; and (c) the mass flow rate, temperature, and pressure at the test section inlets and outlets. These primary test data were used in conjunction with a standard data reduction methodology to estimate essential thermal parameters such as heat fluxes, heat transfer coefficients, and condensation rates. The effects of NCGs on steam condensation within the geometry of the scaled test sections were then presented in regard to various testing conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Machine-Learning Analysis of Moisture Carryover in Boiling Water Reactors

Moisture carryover (MCO) is modeled in the General Electric Type-4 boiling water reactor (BWR) using machine-learning methods and data from operating plants. Understanding MCO and the conditions that give rise to an elevated value is important since excessive MCO can damage critical turbine components, can result in elevated dose levels to on-site personnel, and can interfere with late-cycle power management. The analysis of MCO takes into account simplifying reactor symmetries and important geometric dependencies. The plant data are taken from several reactors and were collected over multiple years and multiple fuel cycles. A brief description of the origin of MCO in U.S. BWR plants is given. A machine-learning model is constructed from the data using applicable algorithms and data-reduction techniques. Matching model complexity with available data is one of the more challenging machine-learning tasks. Too many features and too little data will lead to overfitting. The data for each fuel cycle included over 6876 original features, 9 for each fuel bundle. Two approaches are used to reduce the data set into a manageable number of features. The first was an engineering analysis that resulted in the selection of steam quality Q and steam liquid phase velocity V L as the main features driving MCO. Using a Q and a V L for each fuel bundle gives 1528 Q and a V L feature describing the reactor behavior. An analysis of different functional forms of these two variables led to the actual inputs to the neural network model. The second approach involved the use of statistical techniques such as Pearson’s correlation and k-means analysis. The identified groupings of bundles behaved similarly. Treating each grouping as a single feature further reduced the input variable set to a manageable number. Here, a model selection criterion is proposed, and results are presented along with a discussion of related issues.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1976

Comparative performance of solar thermal power generation concepts

A performance comparison is made between the central receiver system (power tower) and a distributed system using either dishes or troughs and lines to transport fluids to the power station. These systems were analyzed at a rated capacity of 30 MW of thermal energy delivered in the form of superheated steam at 538 C (1000 F) and 68 atm (1000 psia), using consistent weather data, collector surface waviness, pointing error, and electric conversion efficiency. The comparisons include technical considerations for component requirements, land utilization, and annual thermal energy collection rates. The relative merits of different representative systems are dependent upon the overall conversion as expressed in the form of performance factors in this paper. These factors are essentially indices of the relative performance effectiveness for different concepts based upon unit collector area. These performance factors enable further economic tradeoff studies of systems to be made by comparing them with projected production costs for these systems.

Wen, L.↗
DOE OSTI2022

Enhancing reactant selectivity for Ni/Mg reforming catalysts using silicalite-1 shells: A modeling study

The ability of a zeolite shell to enhance the selective conversion of hydrocarbons through diffusional limitations was investigated using a multi-scale model of a fixed-bed reactor. The impact of shell thickness and molecule/pore size on the catalytic performance of silicalite-1 zeolite encapsulated nickel catalyst pellets for steam reforming of C1-C7 hydrocarbons is reported. A reaction–diffusion model using kinetic expressions established in literature was employed. The model was verified through comparison with reported experimental results for steam reforming data over a temperature range of 748 – 1113 K and pressure of 1 – 10 bar. Comparisons are also made against experimental data for steam reforming in the presence of a zeolite shell. Evaluation of the Weisz-Prater criterion for both the core and encapsulated catalyst confirmed mass transfer limitation induced by the utilization of a zeolite shell. The model was used to suggest an optimal thickness that balances diffusional limitations imposed by the zeolite layer on methane versus that of the heavier hydrocarbons. The optimum thickness varied as a function of hydrocarbon size and shape which determined the diffusion rates. For toluene and heptane, a 50 nm thick shell was sufficient to wholly prevent reaction. Hydrocarbons like propane and butane required a shell 7.5 and 5 μm thick. Increasing the gas-hourly-space-velocity from 10,000 to 60,000 h –1 caused a decrease in the optimum shell thickness. Furthermore, this approach can be modified for application to other mixed hydrocarbon systems to predict optimal catalyst design.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2020

A data-driven model for thermodynamic properties of a steam generator under cycling operation

The varying electricity demand from coal power plants due to the intermittent nature of renewable sources leads to load-follow and on/off operations referred to as cycling. Cycling causes transients of properties such as pressure and temperature within various components of the steam generation system.These transients cause increased damage because of fatigue and creep-fatigue interactions shortening the life of components. An algorithm is developed to identify cycling operations from the gross power data. The data-driven model based on artificial neural networks (ANN) is developed using 10 years data from Coal Creek Station power plant located in North Dakota, USA to estimate properties of the steam generator components during cycling operations. Furthermore, the uniqueness of this model is the ability to predict component properties for the cycling as well as base-load operations and is reported for the first time. The ANN model estimates the component properties, for a given gross power profile and initial conditions, as they vary during cycling operations. As a representative example, the ANN estimates are presented for the superheater outlet pressure, reheater inlet temperature, and flue gas temperature at the air heater inlet. The changes in these variables as a function of the gross power over the time duration are compared with measurements to assess the predictive capability of the model. Mean square errors of 4.49E-04 for superheater outlet pressure, 1.62E-03 for reheater inlet temperature, and 4.14E-04 for flue gas temperature at the air heater inlet were observed.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2016

A Proposed Correlation for Critical Flow Rate of Steam/Water Flow

In the safety analysis of loss-of-coolant-accident (LOCA) and steam generator tube rupture (SGTR) scenarios of light water reactors, a modeling of a break is very important to predict the accident's result. For example, the design of a break simulator to simulate a LOCA in light water reactors requires accurate knowledge of the leakage flow through the break, whose shape can be assumed as an orifice or a nozzle according to the characteristics of the break shape. In the case of a small-break LOCA (SBLOCA) scenario, the critical flow mostly occurs by sub-cooled and saturated liquid water including a two-phase water flow at relatively high-pressure conditions. Among previous critical flow maps, Moody's maps are well known for their critical flow rate and pressure of a single component and two-phase mixture with respect to the upstream enthalpy and pressure. In Moody's model, a slip ratio defined by maximizing the critical flow rate with respect to the slip ratio was introduced, and thus its predictions tended to be too conservative. Later, Moody suggested other maps based on a homogeneous equilibrium model (HEM), and they were useful to predict a choked condition for a given upstream condition. However, their predictions of the critical flow rate seemed to show quite an underestimation compared to various test data. The author suggested more practicable critical flow maps for a critical mass flux and pressure of a steam-water flow with respect to the upstream stagnation conditions using an extended Henry-Fauske model. From a comparison with the selected test data, suggested maps showed conservative predictions within reasonable ranges. In this report, a new correlation for the critical flow rate was developed to simulate the suggested critical flow maps within a reasonable range of error. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Primary System Test Facility Review

Idaho National Laboratory (INL) is designing a Dynamic Energy Transport and Integration Laboratory (DETAIL) to demonstrate the operation of an integrated energy system as part of supporting research and development needs for the DOE Office of Nuclear Energy Integrated Energy Systems (IES) Program. DETAIL will demonstrate real-time integration with the electrical grid, renewable energy inputs, thermal and electrical energy storage, and energy delivery to an end user. A Thermal Energy Distribution System (TEDS) is being constructed within the DETAIL facility to demonstrate the operation of integrated energy system heat transfer components, thermal energy storage, distribution systems, and instrumentation and control (I&C), which will allow the facility to emulate the distribution of thermal energy to the generation of electrical power, and/or producing non-electrical commodities (e.g., hydrogen). As is the case for other subsystems within DETAIL, TEDS is designed to operate either independently or as a part of an integrated system to demonstrate the operation of integrated components, develop and validate thermal energy transport models and control systems, and study thermal energy inertia and energy storage. In an effort to completely understand how TEDS will integrate with operational plants, data from tests previously run at existing integral system test facilities (IST) will be used to drive the TEDS heater. Alternatively, a real-time data connection would allow steam conditions at the IST to drive the TEDS heater at the same time as having TEDS conditions provide feedback to the IST in terms of steam demand and/or pressure. This integrated testing approach will assess the ability of TEDS to operate during all plant conditions. Three existing integral system test facilities have been reviewed to determine their suitability for demonstrating TEDS capabilities when integrated within the DETAIL. This report summarizes the capabilities of each IST facility and any challenges that must be overcome to meet the operational needs of TEDS and DETAIL.

25 ENERGY STORAGE↗
DOE OSTI2015

Conductance Steamflow relationship

These histograms represent our calibration of conductance of a volcanic geothermal field (with a clay cap) and the observed steam flow rates. See the following paper for further description: Trainor-Guitton, Hoversten,Nordquist, Intani, Value of information analysis using geothermal field data: accounting for multiple interpretations & determining new drilling locations. SEG Abstracts 2015.

15 GEOTHERMAL ENERGY↗
NASA NTRS2011

In-Service Monitoring of Steam Pipe Systems at High Temperatures

An effective, in-service health monitoring system is needed to track water condensation in real time through the walls of steam pipes. The system is required to measure the height of the condensed water from outside the pipe, while operating at temperatures that are as high as 250 C. The system needs to account for the effects of water flow and cavitation. In addition, it is desired that the system does not require perforating the pipes and thereby reducing the structural integrity. Generally, steam pipes are used as part of the district heating system carrying steam from central power stations under the streets to heat, cool, or supply power to high-rise buildings and businesses. This system uses ultrasonic waves in pulse-echo and acquires reflected signal data. Via autocorrelation, it determines the water height while eliminating the effect of noise and multiple reflections from the wall of the pipe. The system performs nondestructive monitoring through the walls of steam pipes, and automatically measures the height of condensed water while operating at the high-temperature conditions of 250 C. For this purpose, the ultrasonic pulse-echo method is used where the time-of-flight of the wave reflections inside the water are measured, and it is multiplied by the wave velocity to determine the height. The pulse-echo test consists of emitting ultrasonic wave pulses from a piezoelectric transducer and receiving the reflections from the top and bottom of the condensed water. A single transducer is used as a transmitter as well as the receiver of the ultrasonic waves. To obtain high resolution, a broadband transducer is used and the frequency can be in the range of 2.25 to 10 MHz, providing sharp pulses in the time domain allowing for higher resolution in identifying the individual reflections.

Bar-Cohen, Yoseph↗
DOE OSTI2016

Air-water and steam-water onset of flooding data at variable pressure in a large diameter tube

An experimental investigation to acquire flooding data for steam-water and air-water fluid pairs in a large-diameter vertical tube at above-atmospheric pressures is being conducted. The experimental program presented herein is the first to use a single test section with controlled boundary conditions for these pairs in a large-diameter vertical tube, enabling consistent comparisons of flooding characteristics. Many data exist for air-water flooding in annular flow within a large-diameter vertical tube and at various pressures. Steam-water flows are of interest for reactor applications but insufficient data are available to enable reliable analysis. Evaluation of the effect of the fluid pair is performed to understand the applicability of the air-water data to steam-water applications. The pressure range for the tests reported herein is 0.1-0.4 MPa (3-45 psig) for the air-water experiments and 0.1-0.2 MPa (3-15 psig) for the steam-water experiments. The steam-water tests were conducted with minimal subcooling to reduce the effect of phase change and permit a more level comparison with the air-water data. All data were taken at the onset of flooding, that is, conditions when flow reversal of the liquid phase first occurs. The data at higher pressures suggest that the momentum transfer necessary for the onset of flooding is achievable with lower superficial gas velocities as the gas density increases. Flooding characteristics are presented in terms of the Kutateladze parameter. If condensation is accounted for, the saturated steam-water data at 0.1 and 0.2 MPa (3 and 15 psig) closely trend with the air-water flooding data. (authors)

42 ENGINEERING↗
DOE OSTI2017

Pilot-Scale Silicone Process for Low-Cost Carbon Dioxide Capture. Final Scientific/Technical Report

GE Global Research has developed, over the last 8 years, a platform of cost effective CO 2 capture technologies based on a non-aqueous aminosilicone solvent (GAP-1m). As demonstrated in a previous funded DOE project (DE-FE0007502), the GAP-1m solvent has increased CO 2 working capacity, lower volatility and corrosivity than the benchmark aqueous amine technology. The current report describes the cooperative program between GE Global Research (GE GRC), and the National Carbon Capture Center (NCCC) to design, construct, and operate a pilot-scale process using GAP-1m solvent to demonstrate its performance at 0.5 MWe. (i) Performance of the GAP-1m solvent was demonstrated in a 0.5 MWe pilot with real flue gas for over 900 hrs. of operation using two alternative desorption designs: a Continuous Stirred Tank Reactor (CSTR), and a Steam Stripper Column (SSC). The CSTR is a one-stage separation unit with reduced space requirements, and capital cost. The alternative is a multi-stage separation column, with improved desorption efficiency. Testing the two desorber options allowed us to identify the most cost effective, and space efficient desorber solution. (ii) CSTR Campaign: The CSTR desorber unit was designed, fabricated and integrated with the pilot solvent test unit (PSTU), replacing the PSTU Steam Stripper Column at NCCC. Solvent management and waste water special procedures were implemented to accommodate operation of the non-aqueous solvent in the PSTU. Performance of the GAP-1m solvent with the CSTR was demonstrated for over 500 hrs. while varying temperature of the desorption (230 – 265 oF), solvent circulation rate (GAP-1m : CO 2 (molar) = 1.5 – 4), and flue gas flow rates (0.2 – 0.5 MWe). Solvent carry-over in the CO 2 product was minimized by maintaining water content below 5 wt.%, and desorption pressure at 7 psig. CO 2 capture efficiency achieved was 95% at 0.25 MWe (GAP-1m : CO 2 = 4 (molar), 230 oF desorption), and 65% at 0.5 MWe (GAP-1m : CO 2 (molar) = 1.5, 248 oF). Solvent loss was dominated by thermal degradation of the rich solvent. (iii) Steam Stripper Column Campaign: Higher expected cost of the solvent vs. aqueous amines makes solvent management a top priority to maintain the low cost for the process. During the testing of the GAP-1m solvent with the CSTR, thermal degradation of the rich solvent was found to be the main mechanism in solvent loss. Small amounts of water in the working solution were found to be an effective way to enable steam stripping, thereby lowering desorption temperature, and hence reducing thermal degradation. Steam stripping also increased working capacity by 30% due to a more efficient desorption. The concept was first tested in a glass stripping column (lab scale, GE GRC), optimized in a continuous bench scale system (2 kWe, GE GRC), and demonstrated in a 0.5 MWe PSTU at NCCC. No special system modifications were required to the PSTU to accommodate the testing of the non-aqueous GAP-1 solvent with the regenerator column. SSC was found to be more robust towards solvent entrainment (H 2 O < 35 wt.%). 90 – 95% CO 2 capture efficiency was achieved under stoichiometric conditions at 0.5 MWe (235 oF desorption, 2 psig and 19 wt. % H 2 O). Both CO 2 capture efficiency and specific duty reached optimum conditions at 18 wt.% H 2 O. Low amine degradation (< 0.05 wt.%/day) was recorded over 350 hrs. of operation. Controlled water addition to GAP-1m solvent decreased the desorption temperature, thermal degradation, and improved the CO 2 working capacity due to more efficient absorption and desorption processes. Under these conditions, the GAP-1m solvent exhibited a 25% increased working capacity, and 10% reduction in specific steam duty vs. MEA, at 10 oF lower desorption temperature. (iv) Techno-economic Analysis: The pilot-scale PSTU engineering data were used to update the capture system process models, and the techno-economic analysis was performed for a 550 MW coal fired power plant. The 1st year CO 2 removal cost for the aminosilicone-based carbon-capture process was evaluated at $48/ton CO 2 using the steam stripper column. This is a 20% reduction compared to MEA, primarily due to lower overall capital cost. CO 2 cost using the CSTR desorber is dominated by the economics of the solvent make-up. The steam stripper desorber is the preferred unit operation due to a more efficient desorption, and reduced solvent make-up rate. Further reduction in CO 2 capture cost is expected by lowering the manufacturing cost of the solvent, implementing flowsheet optimization and/or implementing the next generation aminosilicone solvent with improved stability and increased CO 2 working capacity.

0.5 MW Pilot↗