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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 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Power Grid Simulation Testbed for Transactive Energy Management Systems

To effectively engage demand-side and distributed energy resources (DERs) for dynamically maintaining the electric power balance, the challenges of controlling and coordinating building equipment and DERs on a large scale must be overcome. Although several control techniques have been proposed in the literature, a significant obstacle to applying these techniques in practice is having access to an effective testing platform. Performing tests at scale using real equipment is impractical, so simulation offers the only viable route to developmental testing at scales of practical interest. Existing power-grid testbeds are unable to model individual residential end-use devices for developing detailed control formulations for responsive loads and DERs. Furthermore, they cannot simulate the control and communications at subminute timescales. To address these issues, this paper presents a novel power-grid simulation testbed for transactive energy management systems. Detailed models of primary home appliances (e.g., heating and cooling systems, water heaters, photovoltaic panels, energy storage systems) are provided to simulate realistic load behaviors in response to environmental parameters and control commands. The proposed testbed incorporates software as it will be deployed, and enables deployable software to interact with various building equipment models for end-to-end performance evaluation at scale.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2020

Conjugate Heat Transfer Analysis of the Supercritical CO2 Based Counter Flow Compact 3D Heat Exchangers

Compact heat exchangers using supercritical fluids such as CO2 are preferred due to their high heat transfer capacity and smaller footprint. Three-dimensional conjugate forced convection heat transfer analysis was performed on several shell-and-tube counter-flow microchannel heat exchangers. Numerical simulations were conducted to test effect of change in mass flow rate, hydraulic diameter and various cross sections on the heat transfer. Increasing mass flow rate improved heat transfer up to a maximum value and then decreased downstream with increasing turbulence. Maximum heat transfer was obtained for the micro channel with the smallest hydraulic diameter. Amongst the cross sections analyzed (circular, square, circular with radial ribs, and square with radial ribs), the most uniform distribution of temperature and maximum heat transfer were obtained for circular cross section with radial ribs. An optimally efficient operation of such a heat exchanger can be attained by considering these factors during multi-objective constrained optimization of geometric parameters and requirements for additive manufacturing of such compact heat exchangers.

Chitale, Janhavi↗
DOE OSTI2023

Emulation and Adversarial Analysis of EV Charging Networks

In the effort of decarbonization and evolution of the modern electrical grid, electric vehicles (EVs) play a key part to transform the grid. However, due to the rapid adoption of EVs and the demand of the charging infrastructure required to power said EVs, risk of a cyber-attack may impose serious consequences. There is a need to analyze and protect the charging ecosystem infrastructure from cyber threats before it reaches wide-scale deployment. In an effort to secure vehicle to grid (V2G) communications, standardization is necessary for continued reliable system operation. The protocol ISO 15118 outlines controls and practices that should be implemented for secure vehicle to grid (V2G) communications. The standard is gaining momentum for American markets as the demand for EV infrastructure grows. The adoption of ISO 15118 in American markets poses several challenges: the deployment of a public key infrastructure (PKI) as outlined within the standard, interoperability of charging different EVs with chargers from different manufactures using the PKI, and scaling the ecosystem to meet the demand while managing risks. This project was created to understand potential cyber and scaling challenges of PKI for EV infrastructure through utilizing a series of emulated components mapping to what exists in the EV ecosystem today, and the components of the PKI that are under development. The key nodes within the emulation that are under development are: electric vehicle (EV), electric vehicle supply equipment (EVSE), charge network operator (CNO), certificate authority (CA), and online certificate status protocol (OCSP) that must all interact using secure and trusted communications. With these emulated components and utilizing orchestration methods to rapidly deploy and scale the components, the ability to analyze risks of the ecosystem and address gaps before the PKI ecosystem is fully deployed to production should yield a more robust and mature production charging infrastructure. Our approach will use a modular architecture of virtual machines within an orchestration platform and will target scales of 100s, 1000s, and 10,000s of entities interacting. The core research questions trying to be answered with this scope of work are: what are the impacts of a rogue CA, what are the risks of certificate revocation list (CRL) management, what is the value of OCSP stapling, what components are vulnerable to DOS attacks, and what test effective payloads may impact the components.

charging ecosystem↗
DOE OSTI2012

Deuteron Spin Structure function $g_1$ at low $Q^2$

The spin structure function $g_1(x,Q^2)$ and its moments provide crucial information on the internal structure of the nucleon. At low momentum transfer $Q^2$, one can study the transition from partonic (quark-gluon) to hadronic (nucleonic) degrees of freedom and test effective theories based on QCD, for instance Chiral Perturbation Theory (ChPT). As $Q^2$ goes to zero, the first moment of $g_1$ is constrained by the GDH sum rule and its ChPT extensions, which makes measurements of $g_1$ in this region uniquely interesting. As part of the large program of spin structure function measurements with CLAS at Jefferson Lab, the EG4 experiment measured the cross section difference between electron beam and proton/deuteron target spins parallel and antiparallel to each other (and the beam direction) down to small scattering angles (approx. 7 degrees). From these differences, $g_1$ can be extracted, with minimal model uncertainties, down to $Q^2$ as low as 0.01 GeV$^2$. We will give a brief overview of the experiment and its analysis, and present first preliminary results on the deuteron spin structure function $g_{1d}(x,Q^2)$.

Adhikari, Krishna↗
DOE OSTI2012

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Quantitative Phenomena Identification and Ranking Table (QPIRT) for Bayesian uncertainty quantification

Propagating parameter uncertainty for a nuclear reactor system code is a challenging problem due to often non-linear system response to the numerous parameters involved and lengthy computational times; issues that compound when a statistical sampling procedure is adopted, since the code must be run many times. The number of parameters sampled must therefore be limited to as few as possible that still accurately characterize the uncertainty in the system response. A Quantitative Phenomena Identification and Ranking Table (QPIRT) was developed to accomplish this goal. The QPIRT consists of two steps: a 'Top-Down' step focusing on identifying the dominant physical phenomena controlling the system response, and a 'Bottom-Up' step which focuses on determining the correlations from those key physical phenomena that significantly contribute to the response uncertainty. The Top-Down step evaluates phenomena using the governing equations of the system code at nominal parameter values, providing a 'fast' screening step. The Bottom-Up step then analyzes the correlations and models for the phenomena identified from the Top-Down step to find which parameters to sample. The QPIRT, through the Top-Down and Bottom-Up steps thus provides a systematic approach to determining the limited set of physically relevant parameters that influence the uncertainty of the system response. This strategy was demonstrated through an application to the RELAP5-based analysis of a PWR Total Loss of main Feedwater Flow (TLOFW) accident, also known as feed and bleed' scenario, . Ultimately, this work is the first component in a larger task of building a calibrated uncertainty propagation framework. The QPIRT is an essential piece because the uncertainty of those selected parameters will be calibrated to data from both Separate and Integral Effect Tests (SETs and IETs). Therefore the system response uncertainty will incorporate the knowledge gained from the database of past large IETs. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Experimental study on the operational and the cooling performance of the APR+ passive auxiliary feedwater system

The passive auxiliary feedwater system (PAFS) is one of the advanced safety features adopted in the APR+ which is intended to completely replace the conventional active auxiliary feedwater system. The PAFS cools down the steam generator secondary side and eventually removes the decay heat from the reactor core by introducing a natural driving force mechanism; i.e., condensing steam in nearly-horizontal U-tubes submerged inside the passive condensation cooling tank (PCCT). With an aim of validating the cooling and operational performance of the PAFS, the separate effect test, PASCAL (PAFS Condensing Heat Removal Assessment Loop), is being performed to experimentally investigate the condensation heat transfer and natural convection phenomena in the PAFS. A single nearly-horizontal U-tube whose dimension is same as the prototypic U-tube of the APR+ PAFS is simulated in the PASCAL test. By performing the PASCAL test, the major thermal-hydraulic parameters such as local/overall heat transfer coefficients, fluid temperature inside the tube, wall temperature of the tube, and pool temperature distribution in the PCCT were produced not only to evaluate the current condensation heat transfer model but also to present database for the safety analysis related with the PAFS. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Condensation model for the ESBWR passive condensers

In the General Electric's Economic simplified boiling water reactor (GE-ESBWR) the passive containment cooling system (PCCS) plays a major role in containment pressure control in case of an loss of coolant accident. The PCCS condenser must be able to remove sufficient energy from the reactor containment to prevent containment from exceeding its design pressure following a design basis accident. There are three PCCS condensation modes depending on the containment pressurization due to coolant discharge; complete condensation, cyclic venting and flow through mode. The present work reviews the models and presents model predictive capability along with comparison with existing data from separate effects test. The condensation models in thermal hydraulics code RELAP5 are also assessed to examine its application to various flow modes of condensation. The default model in the code predicts complete condensation well, and basically is Nusselt solution. The UCB model predicts through flow well. None of condensation model in RELAP5 predict complete condensation, cyclic venting, and through flow condensation consistently. New condensation correlations are given that accurately predict all three modes of PCCS condensation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

OECD RIA Benchmark Phase II - Towards a better understanding of the RIA fuel modelling

This paper deals with the second phase of the RIA (reactivity initiated accident) fuel rod code benchmark for water-cooled reactors that began in 2014. This second phase has been organized around 2 complementary activities: the first activity is to compare the results of different simulations on simplified cases in order to provide additional bases for understanding the differences in modelling of the concerned phenomena, while the second activity is focused on the assessment of the uncertainty of the results. The detailed specification and comparison of the results from the first activity is presented in this paper. Based on the Phase-I and Phase-II conclusions, some generic recommendations can be made. First, fuel and clad thermomechanical models (with the associated material properties) should be further improved and validated more extensively against a sound RIA database. Secondly, build-up of a comprehensive and robust database consisting of both separate-effect tests and integral tests should be pursued in the short term to reinforce the basis for assessment of both individual model and model integration into codes. Thirdly, an assessment of the uncertainty of fuel thermo-mechanics is of high interest (which is consistent with the second activity of this RIA benchmark Phase-II). Finally, as RIA fuel codes are more and more likely to be used for reactor accident studies, particularly for those involving safety analyses, the fuel rod failure criteria (generally used in such analyses) will have to be carefully justified and validated. The current RIA fuel failure criteria are mainly based on the fuel thermal variables and the verification is based on 'conservative' assumptions for the heat transfer conditions. As all codes give rather consistent evaluations of such variables, it appears possible, taking into account adequate provisions, to derive criteria based on thermal variables from experimental values or from an analytical approach. However, if in the future more mechanistic modelling is ever to be used to establish fuel-failure criteria based on mechanical variables, weaknesses in the validation of the codes identified above will have to be accounted for.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

TRACE Code Validation of BWR SVEA Fuel Assembly Spray Cooling

Emergency spray cooling injection is a safety feature implemented in many boiling water reactor designs to reflood a reactor during an accident. Significant experimental work has qualified the efficacy of spray cooling, and ongoing computational modeling efforts strive to more accurately portray the phenomena involved. This work examines the physical phenomena pertaining to emergency spray cooling injection over SVEA-type fuel assemblies. BWR emergency core cooling system (ECCS) designs commonly incorporate a Low-Pressure Core Spray (LPCS) system that introduces coolant through spray nozzles directed to the top of the core, and are located in the upper plenum region. This design is effective for LOCA response, particularly during the initial refill and reflood stages of accident response where the goal is limiting the peak cladding temperature rise in the core. The U.S. NRC thermal-hydraulics code TRACE version 5.0 Patch 4 has been chosen to simulate the separate-effect tests performed by ASEA-ATOM. The computational model was evaluated by performing forward uncertainty quantification using Dakota as the analysis tool and code driver on 31 identified parameters from the input model and TRACE physical constitutive models. This paper discusses the sensitivity of the investigated parameters with respect to spray cooling simulation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Convective Heat Transfer in Decay Heat Exchanger of Sodium-cooled Fast Reactor

Sodium-to-sodium heat exchangers such as the decay heat exchanger (DHX) are key components of the prototype generation [V sodium-cooled fast reactor (PGSFR). The DHX transfers the decay heat generated in the reactor core to the decay heat removal loop after a reactor shutdown. Shell-and-tube type with straight tube arrangement has been widely adopted for the sodium-to-sodium heat exchanger of the sodium-cooled fast reactor. Thus liquid metal heat transfer characteristics inside the tube and around the tube bundle have to be fully understood to design the sodium-to-sodium heat exchanger and evaluate its thermal performance accurately. Plenty of studies on the convective heat transfer of the liquid metals were conducted to present the Nusselt number (Nu) correlations with respect to the Peclet number (Pe). However, those correlations still deviated from each other considerably and few of correlations for low Pe range (less than 100) have been reported. Accordingly, it is difficult for a design code developer to choose a proper heat transfer coefficient correlation for the sodium-to-sodium heat exchanger particularly in a low flow rate range. Recently, a separate effect test for the DHX has been completed using the STELLA-1 facility at KAERI according to the STELLA program which aims at establishment of separate and integral test database (DB) for the design of the PGSFR. The heat transfer performance test of the DHX was successfully conducted under major operating conditions. The objective of this work is to suggest proper convective heat transfer correlations for tube- and shell-sides of the sodium-to-sodium heat exchanger based on the test DB obtained from STELLA-1. The Pe range of the analyzed experimental data is concentrated on less then 100. The adequacy of the suggested correlations is evaluated by applying to the DHX design code, SHXSA. Based on the DHX test DB of the STELLA-1, the proper convective heat transfer correlations for the tube- and shell-side of sodium-to-sodium heat exchangers were suggested. Two correlation combinations for each side were proposed and turned out to be reasonable through the evaluation using the SHXSA design code: Suggestion no. 1: - Lyon correlation (constant = 4) for tube-side; - Lubarsky-Kaufman correlation for shell-side. Suggestion no. 2: - Modified Lyon correlation (constant = 1.45) for tube-side; - Schad-modified correlation for shell-side The Pe ranges of the DHX test DB were 53 - 225 for the tube-side and 45 - 148 for the shell-side, but most of test conditions were concentrated on the Pe range of less than 100. Two correlation suggestions can be practically used for the design of sodium-to-sodium heat exchangers. However, by utilizing more test DB, it needs to be verified which one is closer to real phenomena. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Simulation of Proposed Heat Treatment Methods for the Sister Rods

The High Burnup Spent Fuel Data Project (the 'project'), sponsored by the U.S. Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally evaluating the effects of long-term dry storage and transportation on high burnup (HBU) pressurized water reactor (PWR) fuel (>45 gigawatt days per metric ton uranium). The goals of the project are to provide confirmatory data for analytical model validation and potential improvement; to provide input to future dry storage cask design; to support license renewals and new licenses for Independent Spent Fuel Storage Installation facilities; and to support transportation licensing for high burnup used fuel. In support of these project goals, an instrumented TN-32B bolted-lid cask (the Research Project Cask or RPC) at the North Anna reactor site will be loaded with intact, HBU PWR fuel assemblies (Project Assemblies) having four different kinds of cladding: standard Zircaloy-4, low-tin Zircaloy-4, ZIRLO{sup R}, and M5{sup R}. In approximately ten years, the instrumented cask will be opened for inspection and several fuel rods from the Project Assemblies will be characterized. The project has also removed 25 'sister rods' from assemblies at North Anna: 9 directly from the project assemblies and 16 from similarly constructed and operated HBU assemblies. The 25 sister rods are all high burnup and cover the range of domestically-used cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed non-destructive examinations (NDE) followed by destructive examination (DE). Examination of the sister rods will provide a baseline for comparison with the cask rods and will allow for separate effects testing. Dry storage canister loading is completed under water in a designated area of the plant's spent fuel pool. In preparation for dry storage, the fuel assemblies and canister cavity must be drained and dried. Typically, the highest temperatures experienced by the fuel during dry storage are encountered either during drying or transfer of the canister to the storage pad; therefore, as a part of the sister rod testing these temperatures will be simulated in the lab to produce samples that are representative of the peak cladding temperature the fuel experiences either during drying, transfer, or storage. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Study of Melt Formation and Behavior of a Two-Component Debris Bed in LIVE Facility

The main objective of the LIVE program at Karlsruhe Institute of Technology (KIT) is to study the melt pool behavior in the reactor pressure vessel (RPV). The study is carried out in large-scale 2D and 3D facilities and in supporting separate-effects tests in order to provide a reasonable estimate of the remaining uncertainty. Within the LIVE experimental program several tests have been performed with water and with non-eutectic melts (mixture of KNO{sub 3} and NaNO{sub 3}) as simulant fluids to study the heat flux distribution in the conditions when the melt pool is covered by water from the top. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Coupling Microstructural Evolution Simulations to Material Property Degradation Predictions for Plasma-Facing Materials

Reliable material performance is required for plasma-facing material (PFM) candidates. Previous research has shown that plasma and neutron radiation exposure induces microstructural changes in PFMs; changes in thermal and electrical conductivities and in material hardening and embrittlement were also observed after neutron irradiation. These material property changes will negatively impact the performance of the PFMs in a fusion reactor. Despite the well-known connection between material microstructure, properties, and performance, there is a need for validated modeling capabilities connecting PFM property degradation with microstructural evolution under fusion-relevant conditions. We are developing a simulation capability to couple plasma-induced microstructural evolution to material property degradation. Our approach relies on deliberate mapping between individual simulation models and experimental characterization for validation. The open-source Multiphysics Object-Oriented Simulation Environment (MOOSE) software was used for this simulation capability development. A MOOSE phase-field model was coupled with the cluster dynamics code, Xolotl, to predict microstructural evolution. Microstructure characterization techniques, including scanning electron microscopy (SEM), transmission electron microscopy (TEM), and laser scanning confocal microscopy (LSCM) are used to validate these microstructural evolution simulations. Calculation of thermal and electrical conductivities with first principles simulations was performed for bulk material and for grain boundaries; these results are used within MOOSE models to calculate effective thermal and electrical conductivities as a function of grain characteristics. Thermoreflectance and four-probe techniques were employed to measure the thermal and electrical conductivities, respectively. A MOOSE crystal plasticity model was adapted to predict microstructure-sensitive deformation behavior, and X-ray diffraction (XRD) was used to collect bulk dislocation density data for validation. After individual simulation validation, these models are coupled to predict material property changes resulting from plasma exposure. We focused here on an experimental design to emphasize the separate effects of moderate thermal loads and plasma exposure using tungsten. Annealing of tungsten was performed under a protective environment for temperatures ranging from 500 C to 1500 C. The plasma exposure was completed in the Tritium Plasma Experiment at Idaho National Laboratory under a deuterium flux of 1e22 D/m^2-s. This incremental approach is employed to build confidence in the modeling capability: separate-effects tests ensure that the models capture key mechanisms from single environmental conditions before predicting PFM property degradation under combined loads. We will show our early results from coupling these simulation models to predict PFM property changes from microstructural evolution. Comparisons of the simulation results with preliminary validation data will be discussed.

36 - MATERIALS SCIENCE↗
DOE OSTI2026

Connect microstructure evolution to property degradation with validated simulation

Reliable material performance is required for plasma-facing material (PFM) candidates. Previous research has shown that plasma and neutron radiation exposure induces microstructural changes in PFMs; changes in thermal and electrical conductivities and in material hardening and embrittlement were also observed after neutron irradiation. These material property changes will negatively impact the performance of the PFMs in a fusion reactor. Despite the well-known connection between material microstructure, properties, and performance, there is a need for validated modeling capabilities connecting PFM property degradation with microstructural evolution under fusion-relevant conditions. We are developing a simulation capability to couple plasma-induced microstructural evolution to material property degradation. Our approach relies on deliberate mapping between individual simulation models and experimental characterization for validation. The open-source Multiphysics Object-Oriented Simulation Environment (MOOSE) software was used for this simulation capability development. A MOOSE phase-field model was coupled with the cluster dynamics code, Xolotl, to predict microstructural evolution. Microstructure characterization techniques, including scanning electron microscopy (SEM), transmission electron microscopy (TEM), and laser scanning confocal microscopy (LSCM) are used to validate these microstructural evolution simulations. Calculation of thermal and electrical conductivities with first principles simulations was performed for bulk material and for grain boundaries; these results are used within MOOSE models to calculate effective thermal and electrical conductivities as a function of grain characteristics. Thermoreflectance and four-probe techniques were employed to measure the thermal and electrical conductivities, respectively. A MOOSE crystal plasticity model was adapted to predict microstructure-sensitive deformation behavior, and X-ray diffraction (XRD) was used to collect bulk dislocation density data for validation. After individual simulation validation, these models are coupled to predict material property changes resulting from plasma exposure. We focused here on an experimental design to emphasize the separate effects of moderate thermal loads and plasma exposure using tungsten. Annealing of tungsten was performed under a protective environment for temperatures ranging from 500$^o$C to 1500$^o$C. The plasma exposure was completed in the Tritium Plasma Experiment at Idaho National Laboratory under a deuterium flux of 1e22 $\frac{D}{m^2s}$. This incremental approach is employed to build confidence in the modeling capability: separate-effects tests ensure that the models capture key mechanisms from single environmental conditions before predicting PFM property degradation under combined loads. We will show our early results from coupling these simulation models to predict PFM property changes from microstructural evolution. Comparisons of the simulation results with preliminary validation data will be discussed.

36 - MATERIALS SCIENCE↗