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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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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Numerical solution of 2D wet steam flow with non-equilibrium condensation and real thermodynamics

An approach to modeling of wet steam flow with non-equilibrium condensation phenomenon is presented. The first part of our flow model is homogeneous Euler system of transport equations for mass, momentum and total energy of wet steam (mixture). The additional second part describes liquid phase via non-homogeneous system of transport equations for moments of droplets number distribution function and relies on corrected classical nucleation theory. Moment equations are closed by linearization of droplet growth rate model. All necessary relations for thermodynamic properties of steam are provided by IAPWS set of equations. However, properties of condensate are simply modeled by liquid saturation data. Two real equations of state are implemented. Recently developed CFD formulation for entropy (does not require iteration process) and so-called IAPWS special gas equation for Helmholtz energy (one iteration loop is necessary). Flow model is validated on converging-diverging supersonic nozzle with Barschdorff geometry. Simulations were performed by in-house CFD code based on finite volume method and stiff character of equations was solved by symmetrical time operator splitting. Achieved results satisfactorily agreed with experimental data.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2018

U 3 Si 2 behavior in H 2 O: Part I, flowing steam and the effect of hydrogen

Recent interest in U 3 Si 2 as an advanced light water reactor fuel has driven assessment of numerous properties, but characterization of its response to H 2 O environments is absent from the literature. The behavior of U 3 Si 2 in H 2 O containing atmospheres is investigated and presented in a two-part series of articles aimed to understand the degradation mechanism of U 3 Si 2 in H 2 O. Reported here are thermogravimetric data for U 3 Si 2 exposed to flowing steam at 250–470 °C. Additionally the response of U 3 Si 2 to flowing Ar-6% H 2 from 350 to 400 °C is presented. Microstructural degradation is observed following hours of exposure at 350 °C in steam. U 3 Si 2 undergoes pulverization on the timescale of minutes when temperatures are increased above 400 °C. In conclusion, this mechanism is accelerated in flowing Ar-H 2 at the same temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Apparatus and method for acoustic monitoring of steam quality and flow

An apparatus and method for noninvasively monitoring steam quality and flow and in pipes or conduits bearing flowing steam, are described. By measuring the acoustic vibrations generated in steam-carrying conduits by the flowing steam either by direct contact with the pipe or remotely thereto, converting the measured acoustic vibrations into a frequency spectrum characteristic of the natural resonance vibrations of the pipe, and monitoring the amplitude and/or the frequency of one or more chosen resonance frequencies, changes in the steam quality in the pipe are determined. The steam flow rate and the steam quality are inversely related, and changes in the steam flow rate are calculated from changes in the steam quality once suitable calibration curves are obtained.

47 OTHER INSTRUMENTATION↗
DOE OSTI2019

Apparatus and method for acoustic monitoring of steam quality and flow

An apparatus and method for noninvasively monitoring steam quality and flow and in pipes or conduits bearing flowing steam, are described. By measuring the acoustic vibrations generated in steam-carrying conduits by the flowing steam either by direct contact with the pipe or remotely thereto, converting the measured acoustic vibrations into a frequency spectrum characteristic of the natural resonance vibrations of the pipe, and monitoring the amplitude and/or the frequency of one or more chosen resonance frequencies, changes in the steam quality in the pipe are determined. The steam flow rate and the steam quality are inversely related, and changes in the steam flow rate are calculated from changes in the steam quality once suitable calibration curves are obtained.

02 PETROLEUM↗
DOE OSTI2024

Droplet Entrainment in Steam Supply System of Water-Cooled Small Modular Reactors: Experiment and Modeling Approaches

Droplet entrainment in steam-flow is a prominent phenomenon that needs adequate safety and risk analysis of postulated transient and accident scenarios—including experimental investigation and representative modeling and simulation (M&S)—for small modular reactor (SMR) system design and demonstration. This study identifies knowledge gaps by evaluating experimental and computational fluid dynamics modeling approaches to support early-stage reactor system design, testing, and model evaluation. Previous studies reported in the literature for steam-flow entrainment primarily focused on gigawatt capacity pressurized water reactor (PWR) systems. However, entrainment phenomena are even more prominent for PWR-type SMRs due to their more compact integrated designs, which need further research and development. To fill the research gaps, this study provides insight by specifying the phenomena of interest by leveraging the lessons learned from past research, adopting advanced M&S techniques and advanced instrumentation and control. The findings and recommendations are applicable for evaluating steam-flow entrainment models and for designing integral effect test and separate effect test facilities for gaining reactor design approvals.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Effect of disturbances on MED-TVC plant characteristics: Dynamic modeling and simulation

This study describes a detailed steady-state analysis and transient operation of a multi-effect desalination plant with a thermal vapor compressor unit (MED-TVC). The model developed is based on solving the basic conservation equations of mass, energy, and salt. It can predict steady-state operation and transient behavior when the plant is subject to abrupt changes in the main input parameters including motive steam flow rate (m ms ), cooling seawater flow rate (m cw ) and cooling seawater temperature (T cw ). These sudden changes simulate the conditions a MED-TVC plant may be exposed to. Both the steady-state and dynamic aspects of the model are validated against published experimental data. Monitoring the variation of four state variables (vapor temperature, brine level, brine and vapor flow rates) in each effect shows that disturbances in the motive steam flow rate and the cooling seawater temperature have a significant effect on the plant performance. On the other hand, the disturbance in the cooling seawater mass flow rate has only a moderate effect. For all applied disturbances, the change in the brine level is the slowest compared to the changes in vapor temperature, and brine and vapor flow rates. Furthermore, the effect of changing seawater salinity shows a slight effect on the total distillate production and the specific heat transfer area.

42 ENGINEERING↗
DOE OSTI2020

RELAP-7 Application and Enhancement for FLEX Strategies and ATF Behavior under Extended Loss of AC Power Conditions

This report summarizes the results of a three-year research project sponsored by the U.S. Department of Energy (DOE) Nuclear Energy University Program (NEUP) to enhance and apply the RELAP-7 code by adding and improving several important components (e.g., a mechanistic Reactor Core Isolation Cooling (RCIC) system model) for thermal hydraulic studies of LWRs under ELAP conditions and evaluating the time available for transition to portable FLEX equipment. The project team included University of Massachusetts–Lowell, The Ohio State University, Texas A&M University, Idaho National Laboratory and Oak Ridge National Laboratory. In the Fukushima accident, it was found that the RCIC system played a crucial role in delaying core meltdown by almost three days in Fukushima Daiichi Unit 2, because of self-regulated operation of the steam driven RCIC turbine-pump injection system. Steam flow in the convergent-divergent nozzles of the RCIC Terry turbine is two-phase non-equilibrium transonic flow with homogenous nucleation condensation. To more accurately predict the dynamic process and behavior of the transonic compressible steam flow, a one-dimensional transient two-phase analytical model is presented. A simplified four-fluid model was employed in the present work with the consideration of four separate fluid fields: vapor, liquid film, entrained droplets and condensed droplets. The mass, momentum and energy interactions between the fluids were considered and modeled. An extended seven-equation non-equilibrium critical flow model was developed to obtain the critical pressure and velocities of each phase at the nozzle throat. To predict the wetness in the divergent section, a mechanistic nucleation condensation model was integrated in the nozzle analysis model, considering the generation and consequent growth of droplets. The governing differential equations on a staggered grid were discretized using the second-order Lax-Wendroff scheme with a flux limiter, and the Semi-Implicit Method for Pressure-Linked Equation (SIMPLE) algorithm was employed to solve the discrete linear system. To demonstrate the predictability and reliability of the physical models and the numerical method proposed in the present work, three representative nozzles were modeled and simulated. The results show good agreement with the available experimental data, even for condensation shock. Then, the 1D nozzle model was employed to obtain nozzle flow tables of the Terry turbine nozzle for different working pressures which can cover the operation pressure range of the RCIC system. A mechanistic RCIC turbine-pump system model was developed and implemented in the system code TRACE to simulate dynamic responses of the RCIC system under Beyond Design Basis Accident (BDBA) conditions. The turbine-pump governing equations are based on the control volume approach of the angular momentum balance. The physics based mechanistic RCIC model was developed using the TRACE control system components (i.e., signal variables, control blocks, and tables), and incorporated into a TRACE boiling water reactor (BWR) model. The TRACE model in this report has a detailed nodalization of the reactor pressure vessel (RPV), and all of the major flow paths and system components, including the safety relief valves (SRVs) and the containment suppression pool and drywell. Based on the nozzle flow tables generated from the 1D nozzle model developed, the turbine drive torque can be calculated from table lookup. Since the detailed specifications of the RCIC pump are unavailable, the homologous curves for a Bingham pump were used in the current pump component. A station black-out (SBO) accident test problem was selected to demonstrate the TRACE RCIC model. The short-term SBO simulations were performed for two cladding materials: Zircaloy and FeCrAl, to demonstrate the effect of the accident tolerant fuel cladding on fuel heat-up under BDBA accident conditions. The wetwell plays a vital safety role in SBO and other BWR accident scenarios in that it can reduce containment pressure and supply additional core make-up water. The suppression pool temperature distribution has a very large impact on both RPV and containment pressure. Thus, another novel contribution of the project comes mainly from an improved, systems-level wetwell model which can capture buoyancy-induced thermal stratification effects due to steam injection and condensation. A two-zone stratified wetwell model has been implemented in RELAP-7 and some results from that model are presented. This wetwell model is capable of simulating thermal stratification due to a low steam mass injection rate. With a low mass flow rate, the model assumes that all the steam condenses within the pipe and the resulting plume can be approximated with a purely buoyant, heat-source driven model. The wetwell model developed with these assumptions is adequate to simulate slow transients such as extended SBO transients.

42 ENGINEERING↗
DOE OSTI2019

System for the highly autonomous operation of a modular liquid-metal reactor with steam cycle

The invention relates to a nuclear plant in which the power of a nuclear reactor is controlled via demand of a connected electric grid. A naturally circulating nuclear reactor coolant loop is linked to a water/steam loop by means of a steam generator. The water/steam loop consists of an electric power generating unit and a water recirculating and steam control system. The generator is coupled to an external power grid. As power requirements of the grid change, a controller linked to the generator and a three way valve divides steam flow between the expansion turbine and a feedwater heater to boost or retard the power output. Altering the steam flow changes the pressure and temperature in the water/steam system and thus the coolant flow rate. The change in coolant flow allows the reactor core to regulate its reactivity to reach a state of equilibrium to the demand for electric power.

Vilim, Richard B.↗
DOE OSTI2018

Failure Evaluation of Savannah River Site High-Level Waste Evaporator 242-25H - 18480

The Savannah River Site 242-25H High Level Waste Evaporator was designed in the early 1990's, put into service in January 2000, and discovered to be leaking about 0.4 liters per minute in February 2016. In February 2017, remote investigations revealed that the evaporator has multiple sites in the vessel's bottom spherical head that leak when the steam lance is activated. The leakage was contained within the evaporator's cell, which prevented a hazard to personnel and a release to the environment. An evaporator shell leak is unique in the site's nearly six-decade history of high-level waste evaporator operations. The site's evaporators typically have failed by either fatigue or corrosion attack of the thin steam tubes in the tube-bundle, which supplies heat to the liquid waste in the vessel. The 25H Evaporator design emphasized preventing such failures. However, after 16 years of service, the 25H Evaporator vessel shell leaked rather than the steam tubes. The failure evaluation determined that the leak is due to solid particle erosion in the bottom spherical head, induced by steam flow from the lance, which operates about one-third of the time to keep waste mobilized. Over 16 years, the lance steam flow eroded about 5 mm wall thickness over about one-third of the bottom head's internal surface area, which is equivalent to about 5 kg of metal removed. The vessel is being repaired by remotely welding an elliptical head and ring assembly onto the bottom of the existing vessel. Improvements are being considered for future evaporators. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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

Integral Reactor Containment Condensation Model and Experimental Validation

This NEUP funded project, NEUP 12-3630, is for experimental, numerical and analytical studies on high-pressure steam condensation phenomena in a steel containment vessel connected to a water cooling tank, carried out at Oregon State University (OrSU) and the University of Wisconsin at Madison (UW-Madison). In the three years of investigation duration, following the original proposal, the planned tasks have been completed: (1) Performed a scaling study for the full pressure test facility applicable to the reference design for the condensation heat transfer process during design basis accidents (DBAs), modified the existing test facility to route the steady-state secondary steam flow into the high pressure containment for controllable condensation tests, and extended the operations at negative gage pressure conditions (OrSU). (2) Conducted a series of DBA and quasi-steady experiments using the full pressure test facility to provide a reliable high pressure condensation database (OrSU). (3) Analyzed experimental data and evaluated condensation model for the experimental conditions, and predicted the prototypic containment performance under accidental conditions (UW-Madison). A film flow model was developed for the scaling analysis, and the results suggest that the 1/3 scaled test facility covers large portion of laminar film flow, leading to a lower average heat transfer coefficient comparing to the prototypic value. Although it is conservative in reactor safety analysis, the significant reduction of heat transfer coefficient (50%) could under estimate the prototypic condensation heat transfer rate, resulting in inaccurate prediction of the decay heat removal capability. Further investigation is thus needed to quantify the scaling distortion for safety analysis code validation. Experimental investigations were performed in the existing MASLWR test facility at OrST with minor modifications. A total of 13 containment condensation tests were conducted for pressure ranging from 4 to 21 bar with three different static inventories of non-condensable gas. Condensation and heat transfer rates were evaluated employing several methods, notably from measured temperature gradients in the HTP as well as measured condensate formation rates. A detailed mass and energy accounting was used to assess the various measurement methods and to support simplifying assumptions required for the analysis. Condensation heat fluxes and heat transfer coefficients are calculated and presented as a function of pressure to satisfy the objectives of this investigation. The major conclusions for those tests are summarized below: (1) In the steam blow-down tests, the initial condensation heat transfer process involves the heating-up of the containment heat transfer plate. An inverse heat conduction model was developed to capture the rapid transient transfer characteristics, and the analysis method is applicable to SMR safety analysis. (2) The average condensation heat transfer coefficients for different pressure conditions and non-condensable gas mass fractions were obtained from the integral test facility, through the measurements of the heat conduction rate across the containment heat transfer plate, and from the water condensation rates measurement based on the total energy balance equation. 15 (3) The test results using the measured HTP wall temperatures are considerably lower than popular condensation models would predict mainly due to the side wall conduction effects in the existing MASLWR integral test facility. The data revealed the detailed heat transfer characteristics of the model containment, important to the SMR safety analysis and the validation of associated evaluation model. However this approach, unlike separate effect tests, cannot isolate the condensation heat transfer coefficient over the containment wall, and therefore is not suitable for the assessment of the condensation heat transfer coefficient against system pressure and noncondensable gas mass fraction. (4) The average condensation heat transfer coefficients measured from the water condensation rates through energy balance analysis are appropriate, however, with considerable uncertainties due to the heat loss and temperature distribution on the containment wall. With the consideration of the side wall conduction effects, the results indicate that the measured heat transfer coefficients in the tests is about 20% lower than the prediction of Dehbi’s correlation, mainly due to the side wall conduction effects. The investigation also indicates an increase in the condensation heat transfer coefficient at high containment pressure conditions, but the uncertainties invoked with this method appear to be substantial. (5) Non-condensable gas in the tests has little effects on the condensation heat transfer at high elevation measurement ports. It does affect the bottom measurements near the water level position. The results suggest that the heavier non-condensable gas is accumulated in the lower portion of the containment due to stratification in the narrow containment space. The overall effects of the non-condensable gas on the heat transfer process should thus be negligible for tall containments of narrow condensation spaces in most SMR designs. Therefore, the previous correlations with noncondensable gas effects are not appropriate to those small SMR containments due to the very poor mixing of steam and non-condensable gas. The MELCOR simulation results agree with the experimental data reasonably well. However, it is observed that the MELCOR overpredicts the heat flux for all analyzed tests. The MELCOR predicts that the heat fluxes for CCT’s approximately range from 30 to 45 kW/m2 whereas the experimental data (averaged) ranges from about 25 to 40 kW/m2. This may be due to the limited availability of liquid film models included in MELCOR. Also, it is believed that due to complex test geometry, measured temperature gradients across the heat transfer plate may have been underestimated and thus the heat flux had been underestimated. The MELCOR model predicts a film thickness on the order of 100 microns, which agrees very well with film flow model developed in this study for scaling analysis. However, the expected differences in film thicknesses for near vacuum and near atmospheric test conditions are not significant. Further study on the behavior of condensate film is expected to refine the simulation results. Possible refinements include but are not limited to, the followings: CFD simulation focusing on the liquid film behavior and benchmarking with experimental analyses for simpler geometries. 16 1 INTRODUCTION This NEUP funded project, NEUP 12-3630, is for experimental, numerical and analytical studies on high-pressure steam condensation phenomena in a steel containment vessel connected to a water cooling tank, carried out at Oregon State University (OrSU) and the University of Wisconsin at Madison (UW-Madison). The experimental results are employed to validate the containment condensation model in reactor containment system safety analysis code for integral SMRs. Such a containment condensation model is important to demonstrate the adequate cooling. In the three years of investigation, following the original proposal, the following planned tasks have been completed: (1) Performed a scaling study for the full pressure test facility applicable to the reference design for the condensation heat transfer process during design basis accidents (DBAs), modified the existing test facility to route the steady-state secondary steam flow into the high pressure containment for controllable condensation tests, and extended the operations at negative gage pressure conditions (OrSU). (2) Conducted a series of DBA and quasi-steady experiments using the full pressure test facility to provide a reliable high pressure condensation database (OrSU). (3) Analyzed experimental data and evaluated condensation model for the experimental conditions, and predicted the prototypic containment performance under accidental conditions (UW-Madison). The results are applicable to integral Small Modular Reactor (SMR) designs, including NuScale, mPower, Westinghouse SMR, Holtec-160 and other integral reactors with small containments of relatively high pressures under accidental conditions. Testing has been conducted at the OrSU laboratory in the existing MASLWR (Multi-Application Small Light Water Reactor) integral test facility sponsored by the US Department of Energy. Its highpressure stainless steel containment model (~2 MPa) is scaled to the NuScale SMR currently under development at NuScale Power, Inc.. Minor modifications to the model containment have been made to control the non-condensable gas fraction and to utilize the secondary loop stable steam flow for condensation testing. UW-Madison has developed a containment condensation model, which leveraged previous validated containment heat transfer work carried out at UW-Madison, and extended the range of applicability of the model to integral SMR designs that utilize containment vessels of high heat transfer efficiencies. In this final report, the research background and literature survey are presented in Chapter 2 and 3, respectively. The test facility description and modifications are summarized in Chapter 4, and the scaling analysis is introduced in Chapter 5. The tests description, procedures, and data analysis are presented in Chapter 6, while the numerical modeling is presented in Chapter 7, followed by a conclusion section in Chapter 8.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Aftercooler exhaust duct protection

A turbine engine assembly includes a condenser that is at least partially disposed within the core flow path where water is extracted from the exhaust gas flow, an evaporator system that is at least partially disposed within the core flow path that is upstream of the condenser where thermal energy from the exhaust gas flow is utilized to generate a steam flow. An aftercooler provides a cooling flow that is selectively injected into the core flow path upstream of at least the condenser for cooling the exhaust gas flow in response to a parameter that is indicative of an engine operating parameter that exceeds a predefined condition.

Sobanski, Jon Erik↗
DOE OSTI2016

Experimental study on the guillotine break accident of the main steam line with Atlas

A main steam line break (MSLB) accident is defined as a pipe break in the main steam system of a nuclear power plant. The MSLB accident is characterized as cool-down event due to the increased steam flow rate, which causes excessive energy removal from the steam generators and eventually reactor coolant system. This results in a decrease in reactor coolant temperature. The cool-down causes an increase in core reactivity due to the negative moderator and Doppler reactivity coefficients. The thermalhydraulic integral effect test was performed to simulate a double-ended guillotine break accident of the main steam line of a pressurized water reactor. With an aim of simulating the MSLB accident as realistically as possible, a pertinent scaling approach was taken, especially focusing on a break flow. The main objectives of this test were not only to provide physical insight into the system response of a nuclear power plant during the MSLB accident but also to produce an integral effect test data to validate the safety analysis codes. In the present test, asymmetric thermal-hydraulic behavior was clearly observed between the loops. In the cold leg-1A and -1B where the affected steam generator is connected, the flow rates were higher and the fluid temperatures were lower than those in the cold leg-2A and -2B. This could be attributed to the excessive heat removal in the cold leg-1A and -1B through the affected steam generator. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes. Furthermore, this data can be utilized to identify any code deficiency for a MSLB simulation, especially for DVI-adapted plants. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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↗
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↗
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↗