Search NASA⌕ Search

SEARCH · Search NASA

Results for “Steam Generator”

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.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 73 records · Page 4
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Component Test Facility (Comtest) Phase 1 Engineering For 760°C (1400°F) Advanced Ultrasupercritical (A-USC) Steam Generator Development

The Babcock & Wilcox Company (B&W) performed a Pre-Front End Engineering Design (Pre-FEED) of an A-USC steam superheater for a proposed component test program achieving 760°C (1400°F) steam temperature. This would lead to follow-on work in a Phase 2 and Phase 3 that would involve detail design, manufacturing, construction and operation of the ComTest. Phase 1 results have provided the engineering data necessary for proceeding to the next phase of ComTest. The steam generator superheater would subsequently supply the steam to an A-USC prototype intermediate pressure steam turbine. The ComTest program is important in that it will place functioning A-USC components in operation and in coordinated boiler and turbine service. It is also important to introduce the power plant operation and maintenance personnel to the level of skills required and provide the first background experience with hands-on training. The project will provide a means to exercise the complete supply chain events required in order to practice and perfect the process for A-USC power plant design, supply, manufacture, construction, commissioning, operation and maintenance. Representative participants will then be able to transfer knowledge and recommendations to the industry. ComTest is conceived in the manner of using a separate standalone plant facility that will not jeopardize the host facility or suffer from conflicting requirements in the host plant’s mission that could sacrifice the nickel alloy components and not achieve the testing goals. ComTest will utilize smaller quantities of the expensive materials and reduce the risk in the first operational practice for A-USC technology in the United States. Components at suitable scale in ComTest provide more assurance before putting them into practice in the full size A-USC demonstration plant.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2025

Flow instabilities in helical-coil steam generators for small modular reactors: A review

Here, this study covers the research and discoveries in two-phase flow-boiling instabilities available in the literature—specifically for a helical-coil steam generator (HCSG), including experimental findings, theoretical research, computational models, and system code analyses—supporting research and development of representative small modular reactors (SMRs). Like other new and advanced reactor systems, water-cooled SMRs require experimental data from both integral and separate thermal-hydraulics test facilities for the verification and validation (V&V) of the computational models and computer codes in order to design and obtain regulatory approval. The complex dynamics of two-phase flow-boiling instabilities includes flow regimes physics phenomena, flow-channel geometries, heat-transfer behavior, and interactions among the solid–liquid-gas within the system boundary, all of which are pivotal for understanding the design and operational challenges of SMRs. This study focuses on identifying the relevant knowledge gaps on boiling instabilities—specifically for a HCSG—and provides insights about future research direction optimizing the transport of thermal energy, mass-flow rates, and boundary conditions that ensure the adequate heat-transfer performance, operational stability, and safety associated with SMR systems.

20 FOSSIL-FUELED POWER PLANTS↗
DOE OSTI2016

A simulation of a steam generator tube rupture in AP1000 nuclear power plant based on RELAP5-MV

In the operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions with different codes. In the field of nuclear safety, steam generator tube rupture (SGTR) is one of the main accidents. In this research, transient analysis of the primary system variation of thermo-hydraulics parameters in primary loop under SGTR accident in AP1000 nuclear power plant (NPP) is analyzed by Relap5-MV thermo-hydraulics code. The transient test scenario is modeled using the RELAP5-MV for the AP1000 NPP. The calculation was performed up to a transient time of 3,000.0 s. The results obtained from RELAP5 and RELAP5-MV are in agreement as they are based on the same models though in comparison with RELAP5, RELAP5-MV makes simulation of nuclear power systems easier and convenient for users most especially beginners. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Light water reactor with condensing steam generator

A light water reactor for generating power that utilizes circulation of a primary coolant at saturation pressure to cool a nuclear core and transfer heat from the core to a secondary coolant through one or more heat exchangers of a condensing steam generator. The secondary coolant, once heated can drive power generation equipment, such as steam turbines or otherwise, before being condensed and returned to the one or more heat exchangers.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Design of a Test Article and Test Facility to Investigate Flow-Induced Vibration of a Helical Coil Steam Generator

NuScale Power LLC (NuScale) is developing a small modular reactor (SMR) with unique design features and arrangements that require testing to provide quality data for safety analysis codes and design validation, as well as for technology maturation. A multi-phase testing program is underway at NuScale to demonstrate that its unique helical coil steam generator design is not susceptible to flow-induced vibration during operation. The phases of the testing program include modal testing of individual tubes during test article fabrication and modal and flow testing after test article completion. NuScale has contracted SIET S.p.A. in Piacenza, Italy for the design and fabrication of the test article and test facility. In-fabrication testing was performed at the site of test article build in Cremosano, Italy. Modal testing and flow testing of the completed test article will be performed at SIET’s facilities in Piacenza. This paper discusses the design of the test article and test facility, including supporting systems, required to generate the quality data required to support NuScale’s comprehensive vibration assessment program as part of the overall licensing effort for NuScale’s first-of-a-kind small modular light water reactor.

Flow-induced vibration, helical coil, steam genera↗
DOE OSTI2022

Flow structures of the cross-flow over a five-layer helically coiled steam generator geometry

The helically-coiled heat exchanger (HCSG) offers advantages over straight tubes such as compactness in geometry, increased heat transfer coefficients, and capability to absorb thermal expansion. In this study, the time-resolved velocity field data of shell-side crossflow over a five-layer helically coiled steam generator geometry was obtained using Particle Image Velocimetry (PIV) between adjacent rods in four regions with up to 5000 frames per second (5 kHz) at approximately Re u = 3600. The slant rod bundle created different flow patterns in different regions. Proper Orthogonal Decomposition (POD) revealed high-energy-mode flow structures. Due to the confined flow channel geometry, the large flow structures were not advected resulting in no strong POD mode paring. Continuous Wavelet Transform (CWT) visualized the characteristics of flow fluctuation in time and frequency domain simultaneously, which ranged in Strouhal number, S u , between 0.03 and 1.17. In the wake regions of the rods, flows changed in both pattern and magnitude over time. Once a pattern formed, it continued for relatively shorter than straight tube arrays, before changing the pattern again. The histograms of the POD time coefficients presented this multimodal flow characteristics. The current study focuses more on whole flow field analyses in the selected regions rather than local point-wise analyses. Finally, this study revealed that various Strouhal numbers can exist in the helically coiled heat exchanger geometry, which was not observed in straight tube arrays. In particular, Strouhal numbers smaller than 0.1 do not seem to be caused by vortex shedding but seem to be the result of the combination of complex geometry and possible multimodal trends. Based on the current observation, further works would involve intensive instantaneous point-wise analyses on the interactions between important locations such as separation points, stagnation points, shear layers, etc.

42 ENGINEERING↗
DOE OSTI2019

Large Eddy Simulation of the Flow Behavior in a Simplified Helical Coil Steam Generator

Large eddy simulation (LES) is conducted for the flow over the shell side of a helical coil steam generator (HCSG) heat exchanger. Simulations are conducted on a simplified experimental test section that represents a one-column region of the helical coils using half-rods. Although the rods are wall-bounded, the flow still exhibits the turbulent characteristics and fluctuations from vortex shedding that one would expect from crossflow around a cylinder. The spectral element, computational fluid dynamics (CFD) code Nek5000, is used to capture the physics, and the results are compared with particle image velocimetry (PIV) measurements. In order to ensure that the turbulence is resolved, analysis is conducted by using the Taylor length scales and normalized wall distance. Sensitivity to the inlet boundary conditions (BCs) and the spatial discretization for different polynomial order solutions are also studied, finding only minor differences between each case. Pressure drop and velocity statistics show reasonable agreement with PIV. Proper orthogonal decomposition (POD) analysis reveals that the primary modes are similar between experiment and simulation, although the LES predicts higher turbulent kinetic energy than does PIV. Overall, the study establishes the resolution and resources required in order to conduct a high-fidelity simulation over 12 helical rods.

Lai, Jonathan K.↗
DOE OSTI2016

Graphite Dust Deposition on HTGR Steam Generator: Effect of Particle-wall Impact - 18520

Generated by abrasion of fuel spheres, graphite dust is an important issue for the operation and maintenance of high-temperature gas-cooled reactor (HTGR), since the transport of fission product is closely coupled with graphite dusts. AVR experiments on dust specific radioactivity indicate that graphite dust deposition on steam generator (SG) should be of special interest. Most previous simulations on graphite dust deposition assume a hit-and-stick nature for particles, which may overestimate the deposition rate. In this work we uses a critical velocity model to account for particle-wall interaction. Combined with computational fluid dynamics simulation and discrete phase method, the deposition of graphite dust in the upper 22 layers of SG tubes was studied with particle-wall interaction considered. The result shows that the deposition rate is highly overestimated if hit-and-stick is assumed. For particles larger than 5 {mu}m, the difference can be as large as 3 times. Particle deposition rate reaches maximum at about 3 {mu}m. As for the spatial distribution, few particles deposit on the first layer of tubes due to relative low Stokes number while more depositions occur after entering bundle zone, which indicates that the unsteady wake among tubes could be a major mechanism for the deposition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Safety Benefits Assessment for Accident Tolerant Fuels in Consideration of Steam Generator Tube Degradation Using Dynamic Event Tree Analysis

Accident tolerant fuel (ATF) is expected to delay or prevent core damage by providing additional coping time under accidents involving loss of core cooling. The effect of extended coping time may vary depending on the plant response to accidents. Age-related component degradation that deteriorates plant performance over time could have an impact on the actual advantages of ATF. The potential safety benefits of two near-term ATF candidates, including Cr-coated Zr cladding and FeCrAl cladding, are assessed for a 2-in. loss-of-coolant accident with failed high-pressure safety injection using the dynamic event tree (DET) approach considering possible stress corrosion cracking of steam generator (SG) tubing under aging. The DET approach allows likelihood quantification of accident sequences leading to core damage, including stochastic variation of system response and human actions during accident mitigation. The safety benefits of the selected ATF claddings in terms of additional coping time and the core damage frequency reduction rate under specified accident situations were quantitatively estimated. The results show that the deployment of the two selected ATF claddings is expected to lead to longer coping times and lower core damage frequency due to the wider safety margin to peak cladding temperature they provide. The safety advantages would be greater as SG tube degradation proceeds. Thus, the two ATF candidates would lead to less severe consequences in terms of likelihood of core damage and susceptibility to the SG tube degradation than UO 2 -Zr fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Review and categorization of existing studies on the estimation of probabilistic failure metrics for Reactor Coolant Pressure Boundary piping and steam generator tubes in Nuclear Power Plants

This article presents the first output of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP), titled Methodology for Assessing Pipe Failure Rates in Advanced Water-Cooled Reactors (AWCRs), launched in 2018. This IAEA CRP is aimed at developing a good-practices document for estimating the probabilistic failure metrics of piping in AWCRs. The reliability of piping that comprise the Reactor Coolant Pressure Boundary (RCPB) is important for maintaining safe and stable operations of Nuclear Power Plants (NPPs) because failure of those piping components could lead to undesirable consequences, such as plant shutdown, costly repair, the occurrence of Loss-of-Coolant Accidents (LOCAs) and, possibly, subsequent core damage or large release events. Probabilistic failure metrics (e.g., failure rate, failure frequency, or failure probability) of RCPB components are the key inputs to the Probabilistic Safety/Risk Assessment (PSA/PRA) and risk management of NPPs. The estimation of probabilistic failure metrics, however, is challenging, especially for AWCRs, due to the lack of operating experience. Therefore, as the first step of the IAEA CRP activities, this paper is developed to provide a literature review of the existing studies (from 2000 to April 2019) on the estimation of probabilistic failure metrics for RCPB piping and Steam Generator (SG) tubes of NPPs and to categorize them based on four criteria: (1) explicitness of incorporation of physical failure mechanisms; (2) types of failure characterization; (3) types of physical models for degradation; and (4) explicitness of consideration of maintenance. The existing studies are also analyzed from the viewpoint of the following key aspects: (i) uncertainty analysis, (ii) sensitivity analysis, (iii) validation strategies, and (iv) the areas of applications.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

IP-2 Steam Generator Lower Assembly Packaging, Transport and Disposal - 16510

WMG, Inc. led a team of industry experts in the first of its kind packaging, transportation and disposal of four (4) Steam Generator Lower Assemblies (SGLAs). The project had a number of significant challenges including qualifying the SGLAs as IP-2 packages, implementing roadway improvements and orchestrating the use of three different modes of transportation. In the end, these SGLAs would be the first large components disposed at the WCS Disposal facility in Andrews Texas. The client was re-purposing the SGLA storage mausoleum for 'Fukushima equipment' preparedness so it became necessary to dispose of the assemblies. The WMG team needed to: - Dismantle the mausoleum to allow removal of the SGLAs, - Design, fabricate, deliver and install IP-2 SGLA packaging, and - Transport the packaged SGLAs from the client to the WCS facility for disposal. After qualifying the packaging as IP-2 containers, the SGLAs were loaded onto special equipment for over-the-road transport to the harbor in Kewaunee, WI. Following trans-load, the SGLAs were barged down Lake Michigan through the Mississippi River to the Port of Houston for another trans-load. The SGLAs arrived by rail at WCS' Andrews TX facility. Each transport segment required coordination for transport clearances, proper marking, labeling and placarding for that particular mode of transport as well as coordination with the team involved in moving the large packages from one conveyance to another. The project was completed with zero safety issues, significantly below dose goals, ahead of schedule, and within budget. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Development of Diffusion-type Hydrogen Meters for Steam Generator Leak Detection System of Sodium-cooled Fast Reactors

The goal of this project is to develop and demonstrate diffusion-type hydrogen meters (DTHMs) for hydrogen concentration in cover gas or in molten sodium. Prompt detection of steam-to-sodium leaks of a sodium-heated steam generator (SHSG) is one of the most important economic and safety issues to be addressed in designing and operating a sodium-cooled fast reactor (SFR). A water-to-sodium leak causes a violent exothermic sodium-water reaction (SWR) resulting in local temperature rise (>1,200°C) and produces hydrogen gas and highly corrosive chemicals. The SWR often induces self-enlargement (self-wastage) of the leak and rapid propagation of wastage to adjacent tubes, leading to secondary or multiple tube failures. If it is unmitigated, the hydrogen gas formation could cause a pressure increase in the reactor and may cause a SHSG blowdown, and ultimately result in shutdown of the reactor. In addition, if their effects are unmitigated, the highly corrosive chemicals formed could lead to and accelerate the corrosion problems of the SHSG and sodium piping.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

SAM Process Heat Model Development and Validation: Liquid-HTF Trough and Direct Steam Generation Linear Focus Systems

Two new solar industrial process heat (SIPH) models have been added to the System Advisor Model (SAM) energy simulation tool. The two process heat models were: i) liquid-heat transfer fluid (HTF) parabolic troughs; and ii) direct steam generation (DSG) linear-focus collectors. The DSG model can simulate either DSG parabolic troughs (PT) or DSG Linear Fresnel (LF) fields, based on the optical characteristics supplied. These models help to meet the need for independent, validated tools that reliably simulate the effects of adding a solar field to provide heat for potential IPH applications. During the pre-validation of the liquid-HTF model, operational data was provided and preprocessed by SPF and evaluated at ZAE Bayern. The beta liquid-HTF model provided excellent agreement between the simulations and the operational data for the solar thermal yield generation on a yearly and monthly basis (4.4% difference). The final liquid-HTF process heat model has been further optimized with test users such as ZAE Bayern and SPF. Key improvements in the SIPH models to better model SIPH fields include the use of smaller piping systems and heat capacities, a single collector field, and a higher time-step resolution than for electricity generation (down to 10 minutes). Both the final liquid-HTF and DSG models were successfully validated against operational data for respective solar fields, and showed less than 5% deviation to actual results. A brief application where the liquid-HTF process heat model is used for a brewery in California and the value of reliably simulating a potential SIPH solar field are highlighted.

14 SOLAR ENERGY↗
DOE OSTI2019

Flexible Design for Steam Generation in an SMR Module

The intermediate heat exchanger is a key aspect of the JUMP tertiary loop. An IHX will be required as an interface between the main steam side and secondary side of the NuScale Power Module and coupled TES or energy users. This project examines the preliminary design of the IHX to establish a design range for detailed design and explore the relationships between variables.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Using Dynamic Simulation to Evaluate Attemperator Operation in a Natural Gas Combined Cycle With Duct Burners in the Heat Recovery Steam Generator

A generic training simulator of a natural gas combined cycle was modified to match operations at a real plant. The objective was to use the simulator to analyze cycling operations of the plant. Initial operation of the simulator revealed the potential for saturation conditions in the final high pressure superheater as the attemperator tried to control temperature at the superheater outlet during gas turbine loading and unloading. Subsequent plant operational data confirmed simulation results. Multiple simulations were performed during loading and unloading of the gas turbine to determine operational strategies that prevented saturation and increased the approach to saturation temperature. The solutions included changes to the attemperator temperature control setpoints and strategic control of the steam turbine inlet pressure control valve.

20 FOSSIL-FUELED POWER PLANTS↗