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At least 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Life Extension of Nuclear Plant Facilities through Monitoring and Detailed Analysis - 16175

The UK has accumulated a substantial legacy of radioactive waste from various civil nuclear and defence programmes. Presently, however most of the UK radioactive waste is generated from nine power stations which generate approximately one sixth of the of the country's electricity. Irradiated fuel is removed from these reactors and safe management of the radioactive solid and liquid effluents is required. The safe management of these effluents is essential. Sellafield, in West Cumbria UK, contains two reprocessing plants, one designed to process nuclear fuel from Magnox reactors (CO{sub 2} cooled with Uranium bar fuel) and the other to process nuclear fuel from Advanced Gas cooled Reactors and Pressurized Water Reactors. The reprocessing stage enables extraction of fissile material for recycling. This leaves a volume of dilute but high-level waste. The high-level radioactive liquid waste is concentrated to reduce its volume and then incorporated into borosilicate glass, using a process called vitrification. The glass is then poured into stainless steel containers and placed in long term storage. This required concentration of HLW liquor is achieved using a suite of aging evaporators. All evaporators are of similar design constructed from corrosion resistant stainless steel having dimensions height ∼11 m, internal diameter 3.05 m and capable of holding 13.7 m{sup 3} of liquor. The basic principle of operation is to heat the dilute liquor under partial vacuum, at constant volume, until the target concentration is achieved. Heating of the liquor is achieved using steam heated coils within the evaporator, and steam in the jacket which surrounds the lower half of the evaporator vessel. The evaporator is then brought to atmospheric pressure and the concentrated liquor is ejected whilst hot. The evaporator is then cooled with water cooled coils and a water filled jacket. New dilute liquor is then added for a new batch. The process is then repeated. As the evaporators reach the end of their original design life accurate prediction of the current condition of nuclear plant facilities is essential to safely justify continued plant operation by ensuring structural integrity. This poses a significant challenge as many aging UK nuclear facilities were not inherently built with inspection in mind. The potential life limiting components of the HA evaporators at Sellafield are the surfaces that contact the product liquor at elevated temperatures i.e. the steam heated surfaces of the coils within the evaporator vessel and the vessel base and wall surrounded by the steam jacket. Each heating component is critical to the operational life of the evaporator - the removal from service of a coil heating component significantly reduces the operational flexibility and remnant life of this plant. Over the last decade a variety of cutting-edge expertise and methodologies have been utilised to accurately measure and then predict the current condition of the plant and also to predict its remnant life. This paper provides an overview of the development of several techniques deployed, namely non-destructive testing/inspection, in-house software development and finite element modelling. These methods have been validated by lab-scale or full-scale experiments along with historical plant data. These techniques have been evolved and refined for over a decade, providing vital insights into the physical behaviour of this complex reprocessing system. This enables not only accurate assessment of the current condition of the plant equipment but also allows scientifically underpinned predictions to be made of how it will behave into the future. This is critical for Sellafield as this knowledge feeds into their safety case to the Office of Nuclear Regulation to justify continued operation. It is concluded that this integrated approach provides a robust analysis tool which has driven change and optimisation of plant procedures ensuring continued safe operation whilst extending the life of the facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Improvements to Modeling Capabilities of ATF Concepts in the BISON Fuel Performance Code

Research into accident tolerant materials for the fuel and cladding in Light Water Reactors has been at the forefront of the nuclear fuel research community since the events that occurred at the Fukushima Daiichi nuclear power plant in March 2011. Accident tolerant materials are defined as those that provide significantly increased response time in the event of an accident while providing similar or improved performance as the conventional UO{sub 2}/zirconium-based cladding fuel rods during normal operation. In particular, qualitative attributes for materials with enhanced accident tolerance include improved reaction (e.g., oxidation) kinetics with steam, resulting in lower hydrogen (or other combustible gases) generation rate, while maintaining acceptable thermo-mechanical properties, fuel-clad interactions, and fission-product behavior. Through it's Office of Nuclear Energy, the United States Department of Energy (U.S. DOE) has accelerated research in this area through the Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC). The goal of the AFC Accident Tolerant Fuel (ATF) program is to guide selection of promising concepts for insertion into a commercial reactor as part of a lead test rod or assembly in 2022. Given the aggressive development schedule, it is impossible to perform a comprehensive set of experiments to provide material characterization data. Therefore, the AFC is utilizing computational analysis tools in an effort to understand the proposed accident tolerant materials. The two materials of interest in this work are iron-chromium-aluminum (FeCrAl) cladding and U{sub 3}Si{sub 2}. From a fuel performance standpoint, the correlations available to describe the behavior of these two materials is limited. Therefore, as additional experimental data becomes available and correlations are updated, the fuel performance models must be updated as well to assure that the material models reflect the latest experimental findings. In this paper we demonstrate the importance of utilizing the latest state-of-the-art models for fuel performance simulations by comparing the change in the updated thermal conductivity model for U{sub 3}Si{sub 2} fuel and oxidation model for FeCrAl cladding alloys using the BISON fuel performance code under normal operating conditions. The results indicate that the latest thermal conductivity model for U{sub 3}Si{sub 2} results in fuel centerline temperatures that are on average approximately 3% higher than the previous model. For the oxidation behavior of FeCrAl alloys the new model predicts observable but small oxide thickness (6 μm) compared to the previous model which predicted essentially zero oxide scale formation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Development of an update to ISA S67.04 and RP 67.04: 'Setpoints for Nuclear Safety-Related Instrumentation for Nuclear Power Plants' - 146

The purpose of this paper is to provide an overview of the update to International Society of Automation (ISA) Standard ISA S67.04 - 2006, 'Setpoints for Nuclear Safety-Related Instrumentation for Nuclear Power Plants', and ISA RP 67.04-2000, 'Methodologies for Determination of Setpoints for Nuclear Safety-Related Instrumentation'. Significant advances have occurred on the utility, Nuclear Steam Supply System (NSSS) vendor, and regulatory side, related to improved techniques for analysis and monitoring. These and other changes, including those related to Technical Specifications and surveillance interval extensions, are being included in the latest update to this primary standard for setpoint control in nuclear plants in the U.S. and around the world. In 2016, the ISA S67.04 Committee agreed to proceed with a revision to this Standard and Recommended Practice, with major updates in the following areas: 1. Definitions - Updates for improved definitions based on TSTF-493, Rev 4, RIS-2006-17 and NRC updated Regulatory Guide (RG) 1.105 2. 95/95 - Updates based on improved guidance on the development of component input data and analysis techniques for 95% probability/95% confidence calculations 3. TSTF-493, RIS 2006-17 updates for Technical Specification input and to support changes 4. NRC input from updates in the Regulatory Guide 1.105 and Branch Technical Position BTP-7-12 revision process. 5. Improved guidance from the Electric Power Research Institute (EPRI) in References 7, 8 and 9. These standards are being updated with a new version anticipated to be issued as a draft in 2017 and submitted for ISA approval in 2019. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Combined Cycle Integrated Thermal Energy Storage “CiTES” (Final Scientific/Technical Report)

The Phase I of this project confirmed the technical feasibility of a Combined Cycle integrated Thermal Energy Storage “CiTES” system, calculated the key performance parameters like power efficiency and costs, and proved its commercial value with full-year simulations for several US electricity markets with high degree of variable renewable generation and volatile hourly electricity prices. The core element of this project is the Electro Thermal Energy Storage (ETES) technology from Siemens Gamesa Renewable Energy GmbH, using thermally stable and inexpensive volcanic rocks as storage material and air as heat transfer medium. This technology is backed by more than 10 years of experience and a 440MMBTU (130MWh-th) pilot plant in Hamburg, Germany, which is in operation since 2019. The integration of this thermal storage in an existing combined cycle power plant (CCPP) is typical power plant technology without any major technology risks. It allows the storage of inexpensive renewable energy during times of surplus renewable generation and the discharge of this energy in times of high energy demand when the fossil plant is in operation. This supplements the fossil power generation with CO2-emission-free energy. The secondary effect of the CiTES system is that a small part of the stored thermal energy is used to keep the heat recovery steam generator (HRSG) and steam turbine (ST) of the combined cycle power plant in hot and ready-to-start condition. This enables the plant to start rapidly when fossil generation is required to satisfy demand as soon variable generation drops off in the evenings or during cloud cover and calm wind periods. Without pre-warming of the HRSG and ST, the CCPP would need several hours for a cold or warm start, burn a lot of gas and release high NOx emissions during start and wouldn’t be able to use the short times of high energy prices in an efficient or economical manner. The economic parameters of CiTES were determined by a full year “8760” simulation using a data set calculation for each of the hours of the year, and historical electricity and gas prices. For consistency, the simulations were focused on the pre-COVID year 2019. The financially most attractive markets were in the Energy Reliability Council of Texas (ERCOT) region, which allowed substantial value generation with arbitrage (charge with cheap energy during renewable surplus times and discharge when energy is needed and expensive). The improvement of flexibility with the CiTES system by pre-heating and warm-keeping of the CCPP allowed for additional power generation during short time periods when demand is high but renewable generation is down; when the hourly energy prices are highest in these markets. The simulations are based on 2019 data, when ERCOT had 27GW of installed photovoltaic (PV) and wind generation. They showed that the created revenue with the prototypically sized CiTES system of 1,000MMBTU (300MWh-th) falls a little bit short of what is expected from a commercially viable investment. The system has optimization opportunities for cost reduction and increased effectiveness which will be realized during a potential Phase II Pre-FEED study following this project. Furthermore, it is safe to assume that a lot of renewable generation capacity will be added all over the US in the coming years. As an example, ERCOT is predicting to more than double its renewable generation from 27GW in 2019 to a forecasted 63GW in 2023. This will increase the amount of renewable overproduction exponentially. This rapid increase of local overproduction and the need to curtail renewable generation is well documented by the California ISO (www.CAISO.com / managing oversupply). However, the simulations also revealed a weakness in the structure of the electricity markets in the US. More specifically, when electricity prices are very low and approaching negative levels, the owners of Variable Renewable Energy (VRE) will curtail a part of their facility to stabilize the price by reducing supply. This results in a situation in which storage facilities, which are integrated in existing fossil assets and don’t have the behind-the-meter benefit of a VRE, won’t be able to purchase low cost – otherwise curtailed – renewable energy off the grid. A special tariff, which motivates VRE owners to sell otherwise curtailed renewable energy to storage facilities (Hydrogen, thermal, pumped hydro, etc.) can solve this issue. The implementation of such a regulating tariff by Independent System Operators, thus avoiding renewable curtailment, is a pre condition for successful commercialization for renewable energy storage technologies. With this advancement of design and technology and improvements in the market environment, it can be expected that the Combined Cycle integrated Thermal Energy Storage proves itself as an important innovation to keep highly efficient, natural gas-based power generation economically successful and relevant for the power industry in the United States of America.

Wolf, Thorsten↗
DOE OSTI2017

Improvements to Modeling Capabilities of ATF Concepts in the BISON Fuel Performance Code

Research into accident tolerant materials for the fuel and cladding in Light Water Reactors has been at the forefront of the nuclear fuel research community since the events that oc- curred at the Fukushima Daiichi nuclear power plant in March 2011. Accident tolerant materials are defined as those that provide significantly increased response time in the event of an accident while providing similar or improved performance as the conventional UO2/zirconium-based cladding fuel rods dur- ing normal operation [1]. In particular, qualitative attributes for materials with enhanced accident tolerance include im- proved reaction (e.g., oxidation) kinetics with steam, resulting in slower hydrogen (or other combustible gases) generation rate, while maintaining acceptable thermo-mechanical prop- erties, fuel-clad interactions, and fission-product behavior [1]. Through its Office of Nuclear Energy the United States De- partment of Energy (U.S. DOE) has accelerated research in this area through the Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC). The goal of the AFC Accident Tolerant Fuel (ATF) program is to guide se- lection of promising concepts for insertion into a commercial reactor as part of a lead test rod or assembly in 2022. Given the aggressive development schedule, it is impossi- ble to perform a comprehensive set of experiments to provide material characterization data. Therefore, the AFC is utilizing computational analysis tools in an effort to understand the proposed accident tolerant materials. The two materials of interest in this work are iron-chromium-aluminum (FeCrAl) cladding and U3Si2. From a fuel performance standpoint, the correlations available to describe the behavior of these two materials is limited. Therefore, as additional experimental data becomes available and correlations are updated, the fuel performance models must be updated to represent the current state of the art understanding of these materials. In this paper we demonstrate the importance of utilizing the latest state-of-the-art models for fuel performance simula- tions by comparing old and new thermal conductivity models for U3Si2 fuel and old and new oxidation models for FeCrAl cladding alloys using the BISON [2, 3] fuel performance code under normal operating conditions. The results indicate that the latest thermal conductivity model for U3Si2 results in fuel centerline temperatures that are on average approximately 3% higher than the previous model. For the oxidation behavior of FeCrAl alloys the new model predicts observable but small ox- ide thickness (~6 µm) compared to the previous model which predicted zero oxide scale formation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Comparing Velocity Profiles Along the Rod Length of a Helical Coil Steam Generator Model

A model of a specific type of tube and shell heat exchanger, the helical coil steam generator (HCSG), was designed and constructed to study multiple fluidic properties that attribute to the efficiency of the heat transfer it provides. Tube and shell heat exchangers are currently used throughout industry in chemical, power and petroleum engineering. Most commonly, a primary flow is circulated through a shell, while the secondary flow is run through several small tubes that are bundled together. Methods to increase the heat transfer between the two flows include adding baffles to force shell side flow over more tubes, fins added in-between layers of tubes and passing tubes multiple times through the shell to increase heat exchanger effectiveness. The HCSG incorporates many of the proposed methods but also relies on a complex geometry to increase the heat transfer between tube and shell side flow. In an area where cost effectiveness is favorable, such as inside a nuclear reactor, helically banked steam generators makes extensive use of the room with no sharp tube bends. This increases the pressure loss on the secondary flow and can solve potential temperature manufacturing issues at the welds of junctions. Additionally, they have exhibited pure cross flow behaviors with high heat transfer coefficients and axial temperature profiles that correspond to pure counter-flow with high mean temperature differences. Early work showed that helical banks can have a significant increase of 16 to 43% higher heat transfer coefficient than straight pipe heat exchangers. While many previous studies have looked at the heat transfer, pressure drop and flow fields of a tube and shell heat exchanger, they all focus on tube bundles that are either in-line or staggered. The early flow visualization studies on these types of bundle arrangements reported that attached eddies were formed alternatively behind each of the rods and were occasionally dragged into the main stream of shell side flow. Later studies suggested that observable periodic vortex shedding was dependent on geometry. Experimental data taken using techniques such as laser Doppler anemometry (LDV) by Simonin and Barcouda (1986) as mentioned in Hassan (2002), hot-wire anemometry with Tsuchida (1982) and particle image velocimetry (PIV) conducted by researchers as late as 2007 have been used to characterize turbulent flow and distinguish flow field structures within in-line and staggered tube bundles. By studying these flow field structures, their contribution to fluidic properties such as heat transfer efficiency and pressure drop can be characterized effectively. Literature review lead to a design that incorporated five layers of coils wrapped around one another. The outermost and middle layers of tubes coiled at a specific pitch in a clockwise direction while the other two layers coiled counterclockwise. The proposed design incorporates a constantly changing lateral pitch ratio, a, between adjacent rods. The constructed test section isolates a single interface between two adjacent coiling rod bundles of twelve rods. Three distinct cross sectional planes along the rod bundles were looked at to study the effect of the local geometry to flow fields, Section A, B and C. Particle Image Velocimetry (PIV) frames of 896 x 848 pixels were captured at each section across a specific height. Profiles of mean velocity fields for each section are compared and discussed as possible features of the changing lateral pitch ratio.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Control of solvent adsorption in porous liquids through solvent-solvent interactions

Type 3 Porous Liquids (PLs) are a new class of materials that offer transformative potential for gas capture and utilization. These PLs are comprised of a bulky solvent with a suspended empty nanoporous material that enables selective and high-capacity capture from dilute and complex gas steams. The relationship between the nanoporous material, the structure of the solvent molecules, and the time dependence of solvent adsorption into the nanoporous material governs long term adsorption properties. Herein, molecular dynamics (MD) calculations evaluated the solvent dynamics of nine neat solvents with a broad range of chemical identities and experimental density measurements probed the time-dependent adsorption of these solvents into the ZIF-8 metal-organic frameworks (MOFs) to form PL dispersions over three weeks. Two of the nine dispersion, using glyceryl triacetate and 2’-hydroxyacetopheneone as solvents, presented sufficiently low solvent infiltration—characterized by solvent adsorption less than 40% of the ZIF-8 pore volume—to be viable as PLs. The experimental solvent adsorption data for ZIF-8 dispersed in the four aromatic solvents (acetophenone, methylbenzoate, 2-isopropylphenol, and 2’-hydroxyacetophenone) was fit to a kinetic model to quantify rate of solvent adsorption into ZIF-8. The rates of adsorption of these four similar-sized solvents demonstrated that solvent adsorption is slowed and limited by solvent clustering dynamics that are, in turn, driven by intermolecular hydrogen bonding. Ultimately, experimental solvent sorption measurements indicated that glyceryl triacetate and 2’-hydroxyacetophenone with ZIF-8 formed PLs with the most stable porosity and the greatest potential for gas capture. In conclusion, the combined experimental-computational materials exploration framework used here reveals a novel relationship between molecular scale solvent dynamics and macroscale solvent adsorption kinetics critical for the discovery and rapid evaluation of Type 3 PL compositions.

Hydrogen bonding↗
DOE OSTI2018

Advanced Ultrasupercritical (AUSC) Materials Thick-Walled Cycling Header Development for ComTest-AUSC (Final Report)

Alstom Power Inc., a wholly owned subsidiary of the General Electric Company (GE) recently completed “Advanced Ultrasupercritical (AUSC) Materials Thick-Walled Cycling Header Development for ComTest-AUSC” under Award No. DE-FE0026183 for the U.S. Department of Energy National Energy Technology Laboratory (DOE NETL). This project developed a design for testing of a high temperature nickel-based superalloy thick-walled header to simulate cycling and flexible plant operation in new and existing plants. The objective was to simulate the expected thermal stresses of a new AUSC plant or a retrofit header in an existing plant to improve flexible dynamic operation while maintaining component integrity and overall power plant reliability. The high performance superalloy materials that enable higher operating temperatures and pressures can support future transformational power technologies including new AUSC plants at 4250-5000 psi, 730°C/760°C or will serve as a replacement component in a repowered or retrofit USC coal powered cycling plant at 3600 psi, 600°C. These alloys can also enable supercritical CO2 cycles, nuclear power, solar power and more efficient natural gas combined cycle units as targeted by the U.S. DOE/NETL Cross-Cutting Research Technology Program Plan. Recently developed advanced nickel alloys such as Inconel 740H (740H) and Haynes 282 (H282) have been successfully tested on a pilot scale for fireside corrosion using steam loop tubing at AUSC temperatures of 760°C in actual operating coal fired boiler. These high strength materials have been demonstrated their adequacy for use in other high temperature boiler components such as superheater tubes. Thick-walled superheater outlet headers which operate at high temperatures and pressure under severe cycling condition are considered as one of the most critical component in the AUSC coal fired boiler. Over the design life of a plant, these headers are exposed to the most severe fatigue and creep damage accumulation and need to be properly designed for the expected cycling in an AUSC plant with flexible operation. These materials need to demonstrate adequate design life of 20 to 30 years for creep and fatigue conditions. The following tasks were performed: 1. A conceptual design was developed of a thick-walled component representative of typical headers with several tube connections and proper flow rates (~100,000 lbs/hr) suitable for Phase II test at 760°C operation. This was based on a process analysis of expected full scale new and component retrofit designs for AUSC; 2. A multi-physics flow and heat transfer computational fluid dynamics (CFD) analysis of the component for proper layout of the header configuration, number and geometry (size) of tubes required for bypass flow, flow rates through the tubing, and sectional pressure drops was performed. Steam to header metal heat transfer film coefficients in the header section and the tubes during the two stages (half load and full load) of the cycling operation were evaluated and used in the stress analysis; 3. Detailed analytical evaluation of the thick-walled pipe and tube connections for mechanical integrity under cycling conditions was performed. This included transient thermal stress analysis and fatigue usage estimation including the tube-to-pipe weld geometries and weld materials; 4. A layout and design of the thick-walled test component was developed and instrumentation required for a test was identified; 5. Continuum Damage Mechanics (CDM) creep prediction models were extended to 740H and H282 and validated with public data on 740H and H282 for use in creep analysis. In addition, detailed creep analyses were performed separately for the design life of the conceptual AUSC superheater outlet header using these General Electric (GE) creep constitutive models. GE was the prime contractor for the project, which was directed by Principal Investigator Robert Schrecengost and Lead Scientist Dr. B. Reddy Ganta. Other collaborative participants in this project included GE Global Research, GE Boiler Engineering, and a Technical Consultant from EPRI.

01 COAL, LIGNITE, AND PEAT↗
DOE OSTI2022

Sensitivity analysis of small modular reactor mPower with ATHLET

Small modular reactor (SMR) possesses the characteristics of less radioactive source term, simpler structure, modular design and features like passive safety as well as low risk of proliferation. Developed and designed by Babcock and Wilcox, the mPower reactor is one of pronouncing small modular reactors that adopts the integrated design of the pressurized water reactor, steam generator and a built-in internal control rod drive mechanism. In this paper, the code ATHLET (Analysis of fossil-hydraulics of Leaks and Transients) is used to model the mPower reactor. At the same time, the sensitivity analysis of the core structure and geometric parameters of the pipeline loop, as well as the steady-state simulation and post-exothermic simulation are carried out depending on the given scenarios. In addition, the sensitivity analysis of pressure, temperature and mass flow in the secondary loop is performed. The results of this study will mitigate some of the frequent misunderstanding in the field and will help to better understand the design of SMR and provide basic research data for further research. However, there are limitations in the current research. Considering that the mPower is still in the conceptual design, the established model is only based on the limited public data while detailed parameters are still in myth, which leads to certain level of uncertainties in the calculation. Besides, the research is mainly on the influence of sensitivity analysis under steady state conditions. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Development of Cr-coated zirconium alloy cladding for enhanced accident tolerance

In the context of the development of Enhanced Accident Tolerant Fuels (EATF), AREVA NP, along with CEA and EDF, is developing, among others, a chromium (Cr) coating (5- 20 μm thick) on zirconium alloy cladding. The coating is deposited on the cladding surface by Physical Vapor Deposition (PVD) and forms a dense and adherent protective layer. It preserves the good mechanical behavior of the current cladding in normal operating conditions while significantly improving the oxidation behavior in accidental conditions. This behavior is being evaluated through out-of-pile corrosion and mechanical tests at the AREVA NP Technical and Research Centers. The corrosion behavior was evaluated through autoclave tests in both PWR water chemistry at 360 and in 415 Celsius degrees steam to increase the corrosion kinetics. In both cases very small weight gain was measured on the order of 1 mg/dm{sup 2} and remains stable throughout the length of the tests. The mechanical behavior of Cr-coated samples was evaluated through tensile tests on Cr-coated and uncoated M5 tube samples at both room temperature and 400 Celsius degrees. The curves demonstrate that the Cr-coated samples exhibit similar mechanical behavior as uncoated samples, and for all the tests performed, the mechanical properties (yield strength 0.2%, ultimate tensile strength, and total elongation) of the Cr-coated samples all fell within the range of values for uncoated samples. In order to obtain in-pile data as early as possible on this solution, an irradiation program (called IMAGO) in the Goesgen reactor will be performed in the years to come. This will be the first irradiation of EATF solutions in a commercial reactor.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2022

Introducing the 9500 Node Distribution Test System to Support Advanced Power Applications: An Operations-Focused Approach

The 9500 Node Test System is a representative section of distribution power system model developed as a part of the GridAPPS-D™ project, an effort funded by DOE as a part of the Grid Modernization Lab Consortium (GMLC) program. The test system was developed to fulfill a growing need to represent the rapidly evolving state of electric distribution systems by combining elements of legacy infrastructure systems, modern feeder topologies, and an anticipated future with smart grid technologies. It also provides a network model capable of supporting the simulation of operational scenarios such as the ones in a utility distribution control center. This test system allows the evaluation of the performance of advanced power applications in real-time, such as one that simulates the operations of an Advanced Distribution Management Systems (ADMS), Distributed Energy Resource Management Systems (DERMS), etc. in a Distribution control center. This model is an extension of the widely used IEEE 8500 Node Test Feeder and is currently being validated to become an IEEE test case to help increase adoption and widespread usage among both academia and industry. It is a full-size model representative of a section of a utility’s distribution system with multiple feeders fed from different substations. The model includes multiple distribution circuits, a sub-transmission system, multiple substations, behind the meter customer rooftop photovoltaics (PV), and multiple utility-scale distributed energy resources. To enable accurate simulations of operational scenarios, the 9500 Node Test System is designed to support procedure-based operations, with the ability to realistically demonstrate switching operations, feeder reconfiguration, adjustment of volt-var control equipment, dispatch of distributed generation, and response to planned and unplanned outages. The 9500 Node Test System includes three radial distribution feeders with 12.3 MVA of average load, consisting of both medium voltage and low voltage equipment each supplied by a different distribution substation. The three distribution feeders are connected to each other through Normally-Open switches which can be closed when needed to simulate restoration scenarios due to a fault. One feeder represents today’s grid with low penetration of customer-side renewables. The second represents a potential future grid with microgrids and 100% renewable penetration. The third has no customer generation resources, a district steam plant, and a utility-scale solar farm. The three diverse circuits were created to allow the simulation of both today’s situation as well as potential future scenarios. All three feeders have customers connected by low-voltage secondary triplex lines. This test system meets all requirements outlined in the report for creating a simulation environment that would enable discussion between key technical stakeholders as well as having the potential to accelerate operational application development and their subsequent testing and integration. The new model is a possible representation of what we believe the distribution grid may look like in the future: a high penetration of renewables, reconfigurable radial and mesh topology, numerous DERs, islanded microgrids, and significantly increased data and measurement density. The system supports both the solution of existing and newer algorithms but also enables the evaluation of applications in a realistic operational environment defined by task oriented procedural steps that represent the interaction between the control center operator and field personnel.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2026

Life Cycle Analysis of Natural Gas Supply Chain and End Use Applications in the United States

Natural gas (NG) plays a crucial role in current and future energy systems in the United States due to its abundance and affordability. In this study a life cycle analysis of the NG supply chain in the United States was conducted using Argonne's R&D GREET model, examining stages from recovery to distribution using reported field data processed and documented by National Energy Technology Laboratory. Supply chain emissions were evaluated across multiple spatial scales, including national average, overall regional production, region-to-region, and basin-to-region scenarios. The GHG intensity of the U.S. average NG supply chain was estimated at 10.3 kg CO 2 e/MMBtu (lower heating value), with a range across regions from 7.8 kg CO 2 e/MMBtu (Northeast) to 15.1 kg CO 2 e/MMBtu (Pacific). The analysis further assessed how upstream NG emissions influence the life cycle GHG emissions of key end-use applications, including electricity generation (0.044–0.086 kg CO 2 e/kWh from upstream NG in combined cycle facilities), hydrogen production (1.04–2.20 kg CO 2 e/kg H 2 for steam methane reforming [SMR] and 1.06–2.23 kg CO 2 e/kg for autothermal reforming [ATR]), and transit bus operation utilizing compressed natural gas fuel (0.19–0.37 kg CO 2 e/mile) and hydrogen fuel (0.12–0.25 kg CO 2 e/mile for hydrogen produced in SMR and ATR).

compression↗
DOE OSTI2020

Filter 'Cook-Off': Filter Selection via Side-by-Side Testing at Actual Process Conditions - 20040

In 2005, the THOR steam reforming fluidized-bed reactor technology was selected by the U.S. Department of Energy for treatment of sodium-bearing waste at the INL. During the final phase of commissioning prior to radiological operations, regeneration of the high temperature filters in the process gas filter (PGF) vessel became ineffective. After approximately 50 days of testing, the differential pressure across the filters began increasing exponentially and the plant had to be shutdown. Subsequent investigation revealed that the Inconel 625 (I625) sintered metal filter media was corroding in the 550 deg. C reducing environment. A pilot plant run tested four candidate replacement filter media. The top candidate was further tested during a long-term pilot plant run to validate the selection. Data from the testing indicate the selected filters will provide reliable service throughout the life of the project, including after in-situ radiological decontamination. This paper presents the data from this effort and reinforces the value of pilot plant testing. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Electric actuator modification on a turbine driven feed pump at Calvert Cliffs nuclear power plant - 296

Feedwater systems are a leading contributor to unit SCRAMs as indicated by the 2015 INPO data, where 9 of the 51 SCRAMs recorded were a result of feedwater issues. Digital feedwater control systems and enhanced feed pump control modifications can greatly improve performance and control of the plant feedwater systems and help drive down the potential for reactivity management events and unit SCRAMs. Westinghouse has a strong background in the implementation of digital feedwater control systems, and in planning for a recently completed feedwater control system upgrade at Calvert Cliffs, the Westinghouse team worked with Calvert Cliffs to design a solution to a number of common operational issues. Primarily, improving feedwater control over the entire operating range, eliminating single points of vulnerability (SPV) in the feedwater/feed pump systems and eliminating operational issues with the feed pump steam admission valve hydraulic control oil system. A robust first-of-a-kind design and implementation program was undertaken to split the existing hydraulic valve gear actuator with separate electric actuators. Westinghouse worked with Calvert Cliffs and Emerson Process Management (EPM) to install and commission General Electric (GE) feed pump turbines with separate electric actuators for the high-pressure (HP) and low-pressure (LP) steam admission valves. This successful mechanical and control modification was a first for both the nuclear and non-nuclear power generation industries. Westinghouse is working with a number of other utilities to implement similar modifications for PWR and BWR feed pump applications. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Impact of humidity on gas transport in polybenzimidazole membranes

Polybenzimidazoles (PBIs) are promising materials for high temperature H 2 /CO 2 separation in applications such as steam reforming and pre-combustion carbon capture where significant amounts of water are often present. However, PBIs are hydrophilic, and the impact of humidity on PBI gas separation properties is relatively unexplored. Furthermore, opportunity exists to elucidate the interplay between plasticization, free volume, and gas transport in glassy polymer membranes such as PBIs. Here, this study investigates the effect of humidity on H 2 , O 2 , and CO 2 permeabilities at 35 °C in a commercial PBI and two sulfone-containing PBIs. Water uptake significantly reduces PBI gas permeabilities at low humidities due to competitive sorption and antiplasticization. At high humidities, plasticization increases the permeabilities of larger gases in more hydrophilic PBIs. Effective fractional free volumes evaluated from gas permeation data and previously reported water sorption and dilation data suggest water plasticizes PBIs by increasing accessible free volume via enhanced molecular dynamics rather than by creating new free volume cavities.

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

Tuning the properties of metal–organic framework nodes as supports of single-site iridium catalysts: node modification by atomic layer deposition of aluminium

The metal–organic framework NU-1000, with Zr 6 -oxo, hydroxo, and aqua nodes, was modified by incorporation of hydroxylated Al( iii ) ions by ALD-like chemistry with [Al(CH 3 ) 2 (iso-propoxide)] 2 followed by steam (ALD = atomic layer deposition). Al ions were installed to the extent of approximately 7 per node. Single-site iridium diethylene complexes were anchored to the nodes of the modified and unmodified MOFs by reaction with Ir(C 2 H 4 ) 2 (acac) (acac = acetylacetonate) and converted to Ir(CO) 2 complexes by treatment with CO. Infrared spectra of these supported complexes show that incorporation of Al weakened the electron donor tendency of the MOF. Correspondingly, the catalytic activity of the initial supported iridium complexes for ethylene hydrogenation increased, as did the selectivity for ethylene dimerization. The results of density functional theory calculations with a simplified model of the nodes incorporating Al( iii ) ions are in qualitative agreement with some catalyst performance data.

catalysis (heterogeneous), materials and chemistry↗
DOE OSTI2015

Fabrication, Welding, and Inspection Techniques in Preparation for In-Reactor Testing of Accident Tolerant Fuels

After the Fukushima events in 2011, DOE-NE in collaboration with nuclear industry shifted R&D emphasis to accident performance of LWR fuels under extended loss of active cooling and steam exposure. DOE-NE has created a roadmap for the “Development of Light Water Reactor Fuels with Enhanced Accident Tolerance.” The mission of the Accident Tolerant Fuel (ATF) Roadmap is to develop the next generation of LWR fuels with improved performance, reliability, and safety characteristics during normal operations and accident conditions and with reduced waste generation. The ultimate goal of the ATF roadmap is to support the insertion of lead fuel rods (LFRs) or lead fuel assemblies (LFAs) of an Accident Tolerant Fuel into a commercial LWR within 10 years (i.e., by the end of FY-2022). As a step toward this goal, an irradiation test series has been developed to assess the performance of proposed ATF concepts under normal LWR operating conditions. Data generated by this test program will be used to establish the feasibility of certain aspects of proposed ATF concepts, as well as provide information to support screening among concepts; as such, it is an integral part of Phase I: Feasibility Assessment and Down-Selection outlined in the ATF Roadmap. This irradiation test series is planned to be performed as a series of drop-in capsule tests to be irradiated in the Advanced Test Reactor (ATR) operated by the Idaho National Laboratory (INL), and it has been designated as the ATF-1 test series. Current fission reactors use zirconium-based fuel cladding because of its extremely low macroscopic thermal neutron absorption cross-section, good high temperature strength, and decent corrosion resistance. However, advanced, innovative materials may provide these same benefits while increasing reactor safety margin, core power density, and fuel utilization. These advanced fuel cladding systems will allow revolutionary cladding performance and enhanced fuel mechanical designs, however, challenges exist in design, analysis and fabrication of innovative, never before tested, fuel cladding systems for in-reactor testing. This paper highlights the challenges associated with design, fabrication and welding, and inspection of innovative materials and actions taken to address those challenges in preparation for the Phase I ATR irradiation testing. The lessons learned from Phase I of this experiment can be used to guide researchers for design and analysis of future in-reactor testing of advanced fuel cladding systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗