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DOE OSTI2017

Preliminary Two-Phase Terry Turbine Nozzle Models for RCIC Off-Design Operation Conditions

This report presents the effort to extend the single-phase analytical Terry turbine model to cover two-phase off-design conditions. The work includes: (1) adding well-established two-phase choking models – the Isentropic Homogenous Equilibrium Model (IHEM) and Moody’s model, and (2) theoretical development and implementation of a two-phase nozzle expansion model. The two choking models provide bounding cases for the two-phase choking mass flow rate. The new two-phase Terry turbine model uses the choking models to calculate the mass flow rate, the critical pressure at the nozzle throat, and steam quality. In the divergent stage, we only consider the vapor phase with a similar model for the single-phase case by assuming that the liquid phase would slip along the wall with a much slower speed and will not contribute the impulse on the rotor. We also modify the stagnation conditions according to two-phase choking conditions at the throat and the cross-section areas for steam flow at the nozzle throat and at the nozzle exit. The new two-phase Terry turbine model was benchmarked with the same steam nozzle test as for the single-phase model. Better agreement with the experimental data is observed than from the single-phase model. We also repeated the Terry turbine nozzle benchmark work against the Sandia CFD simulation results with the two-phase model for the pure steam inlet nozzle case. The RCIC start-up tests were simulated and compared with the single-phase model. Similar results are obtained. Finally, we designed a new RCIC system test case to simulate the self-regulated Terry turbine behavior observed in Fukushima accidents. In this test, a period inlet condition for the steam quality varying from 1 to 0 is applied. For the high quality inlet period, the RCIC system behaves just like the normal operation condition with a high pump injection flow rate and a nominal steam release rate through the turbine, with the net addition of water to the primary system; for the low quality inlet period, the RCIC turbine shaft work dramatically decreases and results in a much reduced pump injection flow rate, and the mixture flow rate through the turbine increases due to the high liquid phase flow rate. The net effect for this period is net removal of coolant from the primary loop. With the periodic addition and removal of coolant to the primary loop, the self-regulation mode of the RCIC system can be maintained for a quite long time. Both the IHEM and Moody’s models generate similar phenomena; however noticeable differences can be observed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

NASA NTRS1987

Desulfurizing Coal With an Alkali Treatment

Experimental coal-desulfurization process uses alkalies and steam in fluidized-bed reactor. With highly volatile, high-sulfur bituminous coal, process removed 98 percent of pyritic sulfur and 47 percent of organic sulfur. Used in coal liquefaction and in production of clean solid fuels and synthetic liquid fuels. Nitrogen or steam flows through bed of coal in reactor. Alkalies react with sulfur, removing it from coal. Nitrogen flow fluidizes bed while heating or cooling; steam is fluidizing medium during reaction.

Ravindram, M.↗

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 NTRS1967

Reaction of steam with molybdenum is studied

Comprehensive report studies the reaction of flowing steam with refractory metals /in particular molybdenum/, in the temperature range of 1100 degrees C. The reaction products are hydrogen gas and molybdenum oxide vapor.

Kilpatrick, M.↗

DOE OSTI2015

Development of LWR Fuels with Enhanced Accident Tolerance

Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U 15 N and U 3 Si 2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U 3 Si 2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U 3 Si 2 will represent a 15% increase in U235 loadings over those in UO₂ fuel pellets. This technology was then applied to manufacture pellets for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U 3 Si 2 /68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO 2 pellets. Pellets and powders of UO 2 , UN, and U 3 Si 2 the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO 2 . Cold spray application of either the amorphous steel or the Ti 2 AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing has been carried out for the SiC/SiC composite/SiC monolith structures. A structure with the monolith on the outside and composite on the inside was developed which is the current baseline structure and a SiC to SiC tube closure approach. Permeability tests and mechanical tests were developed to verify the operation of the SiC cladding. Steam autoclave (420°C), high temperature (1200°C) flowing steam tests and quench tests were carried out with minimal corrosion, mechanical or hermeticity degradation effect on the SiC cladding or end plug closure. However, in-reactor loop tests carried out in the MIT reactor indicated an unacceptable degree of corrosion, likely due to the corrosive effect of radiolysis products which attacked the SiC.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Microstructural and Oxidation Effects of Nb Additions to U3Si2

U3Si2 is a long term, accident-tolerant nuclear fuel candidate for light-water reactors because of its superior thermal conductivity and increased uranium density when compared to traditional uranium dioxide (UO2). While reducing internal thermal stresses and increasing efficiency, U3Si2 exhibits energetic oxidation during certain off-normal and accident scenarios, which include coolant or steam exposure. To mitigate this, Nb is investigated as an alloy constituent to enhance corrosion resistance and increase mechanical strength. The work presented investigates the response of Nb-alloyed U3Si2 to steam atmospheres. A thermogravimetric analysis is conducted in flowing steam to T > 1000 °C to assess oxidation resistance. The phase characterization of as-melted, thermally annealed and post-oxidation compositions with up to 12 vol% Nb by powder X-ray diffraction, scanning electron microscopy, and energy dispersive spectroscopy is reported.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2019

High temperature oxidation and microstructural evolution of cold spray chromium coatings on Zircaloy-4 in steam environments

Oxidation kinetics and microstructural evolution of cold sprayed Cr coatings on Zircaloy-4 at 1130–1310 °C in flowing steam at atmospheric pressure have been studied. The study is aimed at understanding of the response of Cr coated Zr-alloy under a steam environment and high temperatures pertinent to design basis accidents (DBAs) and beyond design basis accidents (BDBAs) in light water reactors (LWRs). Surface morphology, microstructure, and phases of post-oxidation test samples were characterized using Scanning Electron Microscopy (SEM), x-ray diffraction (XRD), and Scanning Transmission Electron Microscopy (STEM). Growth kinetics of the Cr-oxide scale and interdiffusion layers between the Cr coating and the Zr-alloy substrate were quantified from cross-sectional SEM images. Cross-sectional analysis showed that the Cr coatings offered a 50 times reduction in oxidation rate over bare Zircaloy-4 in a 1310 °C steam environment. Oxidation kinetics at 1130 °C followed parabolic law (i.e., n ~ 0.5 in power law kinetics) but at 1230 °C and 1310 °C the value of n was suppressed to below 0.5 possibly due to the volatilization of Cr species at the two highest temperatures. Interdiffusion at coating/substrate interface resulted in formation of a brittle Cr 2 Zr or Zr(Fe, Cr) 2 intermetallic compound layer on the order of micrometer in thickness and scattered Cr-rich precipitates were observed well below the interface within the Zr-alloy substrate after cooling. These experimental results could provide data to LWR system simulation codes for better estimation of coping time in the event of accidents.

36 MATERIALS SCIENCE↗

DOE OSTI2020

Scaling of two-phase water-steam relative permeability and thermal fluxes in porous media

Two-phase water-steam flow conditions are frequently encountered in many engineering applications, including geothermal reservoirs. Although routine calculations are based on the multiphase Darcy’s law, the role of the topology of the flowing phases at the pore-scale is usually neglected in the estimation of relative permeabilities. Instead, the latter are frequently computed using empirical models like the Corey correlation. In this work, we first apply the model for relative permeabilities based on pore-scale flow regimes developed by Picchi and Battiato (2019), Relative permeability scaling from pore-scale flow regimes, Water Resour. Res. 55, 3215–3233, to scenarios typical of geothermal reservoirs and then extend it by deriving the scaling laws for the transmissibilities and the thermal properties as a function of temperature. First, we discuss the scaling behavior of normalized relative permeabilities in terms of viscosity ratio and capillary number of water-steam systems and, then, we provide a validation of the model against experimental data available in the literature. The model captures well the data trends collected in real 3D porous media. These results suggest that water-steam relative permeabilities follow the same scaling behavior of gas-liquid systems where the non-wetting phase is much less viscous than the wetting phase. Finally, we investigate the impact that relative permeabilities have on heat transfer rates at two-phase flow conditions and the scaling of mass and energy transmissibility and thermal properties of the mixture. An estimation of the exergy carried by the two-phase water-steam mixture is also included.

42 ENGINEERING↗

NASA NTRS1990

Aeroacoustic flowmeter

The flowmeter is based on a measurement of phase difference between two points on the circumference of a pipe separated axially by an integral multiple of sound wavelength. Plane sound waves are generated aeroacoustically by a non-protruding ring cavity energized either directly by the flow or by a subsidiary flow of the same medium. The frequency of the aeroacoustic source varies with temperature and therefore the temperature can be obtained. In the case of steam flow, temperature can be measured independently and therefore from the measured frequency (or speed of sound), the quality of wet steam can be measured. The flowmeter is linear in velocity and no calibrations are required.

Shakkottai, Parthasarathy↗

DOE OSTI2016

Derate Mitigation Options for Pulverized Coal Power Plant Carbon Capture Retrofits

Carbon capture and storage (CCS) technologies available in the near-term for pulverized coal-fueled power plants (i.e., post combustion solvent technologies) require substantial capital investment and result in marked decrease in electricity available for sale to the grid. The impact to overall plant economics can be mitigated for new plant designs (where the entire plant can be optimized around the CCS system). However, existing coal-fueled power plants were designed without the knowledge or intent to retrofit a CCS process, and it is simply not possible to re-engineer an existing plant in a manner that it could achieve the same performance as if it was originally designed and optimized for CCS technology. Pairing an auxiliary steam supply to the capture system is a technically feasible option to mitigate the derate resulting from diverting steam away from an existing steam turbine and continuing to run that turbine at steam flow rates and properties outside of the original design specifications. The results of this analysis strongly support the merits of meeting the steam and power requirements for a retrofitted post-combustion solvent based carbon dioxide (CO2) capture system with an auxiliary combined heat and power (CHP) plant rather than robbing the base plant (i.e., diverting steam from the existing steam cycle and electricity from sale to the grid).

01 COAL, LIGNITE, AND PEAT↗

DOE OSTI2018

Effect of input parameters intensity and duration on dynamic performance of MED-TVC plant

Multi-Effect-Desalination (MED) may be exposed to fluctuations (disturbances) in input parameters during operation. Therefore, there is a requirement to analyze the transient behavior of such MED systems. In this work, a dynamic model is developed and used to examine the effect of abrupt and ramp changes in the main operational parameters on the plant behavior and performance. The results show that the disturbance intensity variation has a major role in the desalination plant behavior. For the current MED-TVC configuration, it is recommended to limit the reduction in the seawater cooling flow rate to under 12% of the designed steady-state value to avoid dry out in the evaporators. Here, a reduction in the motive steam flow rate and cooling seawater temperature of more than 20% and 35% of the nominal operating values, respectively, may lead to flooding in the evaporators and a complete plant shutdown. On the other hand, the disturbance period has a minimal effect on plant performance if it avoids the critical values of the disturbance intensity that can cause plant shutdown. Simultaneous combinations of two different disturbances with opposing effects result in a modest effect on the plant operation and they can be used to control and mitigate the flooding/drying effects of the disturbances. For simultaneous combinations of disturbances with similar effect, the plant needs an accurate control system to avoid an operational shutdown.

42 ENGINEERING↗

DOE OSTI2024

Assessments of advanced reactor heat supply to high temperature industrial unit operations: Heat Engines and Heat Pumps

In this report, the feasibility of employing Advanced Nuclear Reactors to supply the necessary high-quality heat to Industrial Processes is investigated. As a first step, the most relevant industrial applications that could benefit from this coupling were identified. Four industries including petroleum refining (distillation, catalyst regeneration, hydro-steam cracking), chemical polymers, steel manufacturing (metals annealing/soaking) and cement manufacturing (calcination of limestone and dolomite industries) were considered. An overview of the heat duties, i.e., the temperatures and thermal powers required by representative plants of the different processes, was provided. Secondly, the thermodynamic conditions of the steam produced by representative designs of Advanced Reactor concepts (Liquid Metal Fast Breeder Reactors, Molten Salt Reactors, and High Temperature Gas-cooled Reactors) were summarized. The comparison between the requirements of the industrial processes and the capabilities of the candidate nuclear power plants showed that only a few processes could benefit from the coupling. Although Advanced Reactors operate at higher temperatures than Light Water Reactors, the thermodynamic conditions of the generated steam flow rates are generally not suitable for the selected applications. At the same time, data indicate that most of the processes can be suitably supplied if the temperature of steam is elevated to 900 °C. One possible solution to boost the quality of the steam from the nuclear island involves the use of a heat pump. Based on a technology similar to that found in refrigerators and air conditioners, heat pumps extract heat from a source, elevate its temperature and transfer it to where it is needed. A simple numerical example illustrating the viability of this method is described. Specifically, the energy conversion cycle of a conventional Pressurized Water Reactor was extended to incorporate a compressor to raise the heating value of the steam produced by the reactor. Two reference configurations, i.e., one where the industrial process coupled to the nuclear power plant exhausts saturated liquid and the other where it exhausts saturated steam, were considered. The thermal efficiencies achievable by increasing the heating value of the reactor steam in this way are significantly greater than by direct electric heating.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

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 OSTI2013

ANTI-PARALLEL EUV FLOWS OBSERVED ALONG ACTIVE REGION FILAMENT THREADS WITH HI-C

Plasma flows within prominences/filaments have been observed for many years and hold valuable clues concerning the mass and energy balance within these structures. Previous observations of these flows primarily come from Hα and cool extreme-ultraviolet (EUV) lines (e.g., 304 Å) where estimates of the size of the prominence threads has been limited by the resolution of the available instrumentation. Evidence of 'counter-steaming' flows has previously been inferred from these cool plasma observations, but now, for the first time, these flows have been directly imaged along fundamental filament threads within the million degree corona (at 193 Å). In this work, we present observations of an AR filament observed with the High-resolution Coronal Imager (Hi-C) that exhibits anti-parallel flows along adjacent filament threads. Complementary data from the Solar Dynamics Observatory (SDO)/Atmospheric Imaging Assembly (AIA) and Helioseismic and Magnetic Imager are presented. The ultra-high spatial and temporal resolution of Hi-C allow the anti-parallel flow velocities to be measured (70-80 km s{sup –1}) and gives an indication of the resolvable thickness of the individual strands (0.''8 ± 0.''1). The temperature of the plasma flows was estimated to be log T (K) = 5.45 ± 0.10 using Emission Measure loci analysis. We find that SDO/AIA cannot clearly observe these anti-parallel flows or measure their velocity or thread width due to its larger pixel size. We suggest that anti-parallel/counter-streaming flows are likely commonplace within all filaments and are currently not observed in EUV due to current instrument spatial resolution.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

NASA NTRS1994

Electromagnetic Probe Technique for Fluid Flow Measurements

The probes described herein, in various configurations, permit the measurement of the volume fraction of two or more fluids flowing through a pipe. Each probe measures the instantaneous relative dielectric constant of the fluid in immediate proximity. As long as separation of the relative dielectric constants of each fluid is possible, several or even many fluids can be measured in the same flow steam. By using multiple probes, the velocity of each fluid can generally be determined as well as the distribution of each constituent in the pipe. The values are determined by statistical computation. There are many potential applications for probes of this type in industry and government. Possible NASA applications include measurements of helium/hydrazine flow during rocket tests at White Sands, liquid/gas flow in hydrogen or oxygen lines in Orbiter engines, and liquid/gaseous Freon flow in zero gravity tests with the KS135 aircraft at JSC. Much interest has been shown recently by the oil industry. In this industry, a good method is needed to measure the fractions of oil, water, and natural gas flowing in a pipeline and the velocity of each. This particular problem involves an extension of what has been developed to date and our plans and program to solve this problem will be discussed herein.

Arndt, G. D.↗

DOE OSTI2014

RELAP-7 Beta Release: Summary of Capabilities

RELAP-7 is a nuclear systems safety analysis code being developed at the Idaho National Laboratory (INL). Building upon the decades of software development at the INL, we began the development of RELAP-7 in 2011 to support the Risk Informed Safety Margins Characterization (RISMC) Pathway. As part of this development, the first lines of RELAP-7 code were committed to the software revision control repository on November 7th, 2011. The overall design goal of RELAP-7 is to take advantage of the previous thirty years of advancements in computer architecture, software design, numerical methods, and physical models in order to provide capabilities needed for the RISMC methodology and to support nuclear power safety analysis. RELAP-7 is built using the INL’s modern scientific software development framework, MOOSE (Multi-physics Object Oriented Simulation Environment). MOOSE provides improved numerical calculations (including higher-order integration in both space and time, yielding converged second-order accuracy). The RELAP-7 code structure is based on multiple physical component models such as pipes, junctions, pumps, etc. Each component can have options for different fluid models such as single- and two-phase flow. This component-based and physics-based software architecture allows RELAP-7 to adopt different physical models for different applications. A relatively new two-phase hydrodynamic model, termed the ''7-Equation model'' for two phasic pressures, velocities, energies, and volumetric fraction, is incorporated into RELAP-7 for liquid-gas (water-steam) flows. This new model allows second-order integration because it is well-posed, which will reduce the numerical error associated with traditional systems analysis codes. In this paper, we provide a RELAP-7 capability list describing analysis features, range of applicability, and reactor components that will be available for the December 15th, 2014 beta release of the software.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2015

Modeling biomass gasification system using multistep kinetics under various oxygen–steam conditions

Biomass gasification is an effective method to convert low‐value lingocellulosic biomass into useful gas products. The ability to predict the gas composition is important to process design. In this study, simulations of biomass gasification were carried out for various feedstocks at different operating conditions, and the model results were validated using experimental data. A multistep kinetics model was developed to predict the syngas composition using different biomass feedstocks under various oxygen‐enriched air and steam conditions. The oxygen level in the gasifying agent ranged from 21% (air) to 45% (pure oxygen). Five different biomass materials were used, including pine, maple–oak mixture, seed corn, corn stover, and switchgrass. The fluidized bed gasifier was maintained at 800°C. The present kinetics model was able to predict the concentrations of H 2 , CO, CO 2 , H 2 O, CH 4 , and N 2 in the syngas, and the predicted compositions were in good agreement with the experimental data. The results show that the model was sensitive to the variation in the oxygen level, steam flow rate, and the composition of the biomass material. As the oxygen level increased, the model also predicted the increase in H 2 , CO, CH 4 , and the H 2 /CO ratio. The mechanism on improving the syngas quality was also explained by analyzing the reaction equations. The increase in hydrogen is believed to be due to the use of steam for gasification. The present model can further be integrated into the process simulation model for predicting the performance of a biorefinery to help utilize biomass energy. © 2015 American Institute of Chemical Engineers Environ Prog, 34: 1148–1155, 2015

Banerjee, Sudhanya↗

DOE OSTI2025

Device for steam cladding oxidation testing at TREAT

To compare the chemical degradation of conventional zirconium alloy (Zry) cladding to advance silicon carbide (SiC) cladding in a post loss of coolant accident (LOCA) environment, new nuclear testing capabilities are necessary. The Transient Reactor Test (TREAT) Facility at Idaho National Laboratory (INL) has matured its transient fuel testing capabilities since its 2017 restart. The most recent experiment architecture is the Transient Water Irradiation System in TREAT (TWIST), which is designed to support qualification of accident tolerant fuels in light water reactors. INL has designed and analyzed a natural circulation steam flow modification for TWIST to produce prototypic conditions of cladding oxidation. The in-situ device will be electrically heated to drive natural circulation. Moreover, the SiC cladding requires heating above 1700 °C to observe failure, thus internal prototypic nuclear heating with radiation effects will be used. Thermal hydraulic analysis with RELAP5-3D (Reactor Excursion and Leak Analysis Program) estimated steam fluxes greater than 50 mg cm −2 s −1 can be achieved. These fluxes are adequate to test Zry cladding according to draft regulatory guides and to test SiC cladding according to past experiments.

Oxidation↗