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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 1,405 records · Page 78
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Optimized Highly Efficient SSCB Using Organic Substrate Packaging for Electric Vehicle Applications

Solid State Circuit Breakers (SSCBs) are an attractive protection solution for their arcless current interruption and fast actuation speeds over mechanical breakers. This paper proposes a Bidirectional SSCB (BSSCB) with a thermally defined and digitally controlled current time profile for fault protection in EV and other low-voltage DC systems. The paper proposes an organic packaging approach utilizing flex circuitry to develop a reliable, cost-effective power module for BSSCBs. The paper studies transient heat transfer in the power modules using finite element analysis (FEA). An RC themal ladder network is extracted to define a fusing curve. To demonstrate and verify the design, a 1kV/50 A SiC MOSFET BSSCB prototype is fabricated and tested, having a power density of 60 W/cm3 and 4x reduction in form factor over presently researched breakers. Also, given are results for 750 V/150 A operation showing interruption in 2.4 μs.

organic power packaging↗
DOE OSTI2017

Intelligent Gate Drive for Fast Switching and Crosstalk Suppression of SiC Devices

This study presents an intelligent gate drive for silicon carbide (SiC) devices to fully utilize their potential of high switching-speed capability in a phase-leg configuration. Based on the SiC device's intrinsic properties, a gate assist circuit consisting of two auxiliary transistors with two diodes is introduced to actively control gate voltages and gate loop impedances of both devices in a phase-leg configuration during different switching transients. Compared to conventional gate drives, the proposed circuit has the capability of accelerating the switching speed of the phase-leg power devices and suppressing the crosstalk to below device limits. Based on Wolfspeed 1200-V SiC MOSFETs, the test results demonstrate the effectiveness of this intelligent gate drive under varying operating conditions. More importantly, the proposed intelligent gate assist circuitry is embedded into a gate drive integrated circuit, offering a simple, compact, and reliable solution for end-users to maximize benefits of SiC devices in actual power electronics applications.

30 DIRECT ENERGY CONVERSION↗
DOE OSTI2022

Methods and apparatus for comprehensive characterization of performance attributes and damage thresholds of ultrafast laser optics

It is presently well understood that the operational performance limits of optics are determined by three fundamental attributes: the initiation of laser-induced damage, the growth of damage sites, and the transient (nondamaging) modification of optical parameters. The comprehensive characterization of the performance limitations of ultrafast optics requires consideration of all three fundamental attributes. Here, the vast majority of literature to date, however, has focused primarily on damage-initiation and testing systems that are largely focused on determining the damage-initiation threshold under single- and multipulse excitation. In this work, we discuss a testing apparatus that was designed to offer the capability to adequately characterize all three of these performance attributes under femtosecond, near-infrared laser irradiation. Key aspects of methodology are discussed, including high-dynamic-range energy control, variable beam size, wavelength tunability, B-integral management, and functional performance characterization to explore the true operational limits of the components. Example results for a metal-dielectric mirror demonstrate the test station’s operation.

42 ENGINEERING↗
DOE OSTI2016

Study of Convection Heat Transfer in a Very High Temperature Reactor Flow Channel: Numerical and Experimental Results

Very High Temperature Reactors (VHTRs) are one of the Generation IV gas-cooled reactor models proposed for implementation in next generation nuclear power plants. A high temperature/pressure test facility for forced and natural circulation experiments has been constructed. This test facility consists of a single flow channel in a 2.7 m (9’) long graphite column equipped with four 2.3kW heaters. Extensive 3D numerical modeling provides a detailed analysis of the thermal-hydraulic behavior under steady-state, transient, and accident scenarios. In addition, forced/mixed convection experiments with air, nitrogen and helium were conducted for inlet Reynolds numbers from 500 to 70,000. Our numerical results were validated with forced convection data displaying maximum percentage errors under 15%, using commercial finite element package, COMSOL Multiphysics. Based on this agreement, important information can be extracted from the model, with regards to the modified radial velocity and property gas profiles. Our work also examines flow laminarization for a full range of Reynolds numbers including laminar, transition and turbulent flow under forced convection and its impact on heat transfer under various scenarios to examine the thermal-hydraulic phenomena that could occur during both normal operation and accident conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Using FIPD and OPTD to Benchmark Metallic Fuel Performance

This report serves as an introduction, tutorial, and benchmark specification for out-of-pile tests on metallic fuel. It introduces a new user to the EBR-II legacy fuel performance test program and the fast reactor fuel performance databases built to preserve the records. It then details the information stored in each database and how to find it. A benchmark specification is included for a small set of out-of-pile tests on U-10Zr fuel to function as a tutorial demonstrating how the legacy fuel performance data sets stored in the FIPD and OPTD databases can be used together to benchmark fuel performance models for steady-state and transient performance.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Operation of Natural Gas Pipeline Networks With Storage Under Transient Flow Conditions

Here, we formulate a nonlinear optimal control problem for intraday operation of a natural gas pipeline network that includes storage reservoirs. The dynamics of compressible gas flow through pipes, compressors, reservoirs, and wells are considered. In particular, a reservoir is modeled as a rigid, hollow container that stores gas under isothermal conditions and uniform density, and a well is modeled as a vertical pipe. For each pipe, flow dynamics are described by a coupled partial differential equation (PDE) system in density and mass flux variables, with momentum dissipation modeled using the Darcy–Wiesbach friction approximation. Compressors are modeled as scaling up the pressure of gas between the inlet and outlet. The governing equations for all network components are spatially discretized and assembled into a nonlinear differential-algebraic equation (DAE) system, which synthesizes above-ground pipeline and subsurface reservoir dynamics into a single reduced-order model. We seek to maximize an objective function that quantifies economic profit and network efficiency subject to the flow equations and inequalities that represent operating limitations. The problem is solved using a primal–dual interior point solver, and the solutions are validated in computational experiments and simulations on several pipeline test networks to demonstrate the effectiveness of the proposed methodology.

03 NATURAL GAS↗
DOE OSTI2022

Simulation of Peach Bottom turbine trip transient with RETRAN-3D using BXFGEN

The Peach Bottom 2 Turbine Trip transient is used to evaluate the capabilities and accuracy of coupled 3-D neutronics/thermal hydraulics best estimate calculations for support of safety analysis. Selected results, obtained with RETRAN-3D, are compared to available measurements and benchmark results. The comparisons are analyzed and the application of RETRAN-3D for a best estimate analysis approach within the safety analysis methodology is discussed. Overall, both the thermal-hydraulic and neutronic results show a good comparison to the experimental data, and the model is considered to accurately predict test data. In particular, the initial acoustic response of the main steam line and the subsequent main pressurization stages are well-predicted. During the main pressurization stage, the timing and magnitude of the in-core local pressurization and associated void compression are properly predicted based upon the agreement between local power range monitors (LPRM) predictions upwards, but the shift is slightly underpredicted. This indicates the methodology is well-posed for simulating this class of transient response.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Spectral emission measurements of lithium on the lithium tokamak experiment

There has been a long-standing collaboration between ORNL and PPPL on edge and boundary layer physics. As part of this collaboration, ORNL has a large role in the instrumentation and interpretation of edge physics in the lithium tokamak experiment (LTX). In particular, a charge exchange recombination spectroscopy (CHERS) diagnostic is being designed and undergoing staged testing on LTX. Here we present results of passively measured lithium emission at 5166.89 A in LTX in anticipation of active spectroscopy measurements, which will be enabled by the installation of a neutral beam in 2013. Preliminary measurements are made in transient LTX plasmas with plasma current, I{sub p} < 70 kA, ohmic heating power, P{sub oh}{approx} 0.3 MW and discharge lifetimes of 10-15 ms. Measurements are made with a short focal length spectrometer and optics similar to the CHERS diagnostics on NSTX [R. E. Bell, Rev. Sci. Instrum. 68(2), 1273-1280 (1997)]. These preliminary measurements suggest that even without the neutral beam for active spectroscopy, there is sufficient passive lithium emission to allow for line-of-sight profile measurements of ion temperature, T{sub i}; toroidal velocity and v{sub t}. Results show peak T{sub i} = 70 eV and peak v{sub t} = 45 km/s were reached 10 ms into the discharge.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Central ion cyclotron emission in the DIII-D tokamak

Collective ion cyclotron emission (ICE) at the ion cyclotron frequency and its harmonics is a potential passive diagnostic of the fast-ion distribution in fusion reactors. ICE is observed in most plasmas in the DIII-D tokamak and is most strongly excited by the fast ions from neutral beam injection. The conventional outboard-edge ICE is detected in H-mode plasmas. However, weaker centrally-localized ICE is measured in L-mode plasmas, including those with negative triangularity shapes. Similar ICE spectra are found with both ICE diagnostics systems, the dedicated magnetic probes and the instrumented antenna straps. Many differences in the behavior of this central ICE are generated by varying the deuterium beam injection angle into deuterium plasmas. The co-current 'near-perpendicular' beam excites the most central ICE from the co-current beams, with this emission detected from the fundamental to the fifth ICE harmonic. However, the counter-current 'near-tangential' beam destabilizes the highest amounts of centrally-localized ICE. This emission is spectrally broader than that driven by the co-current beams and is observed up to its seventh ICE harmonic. The central ICE excited by this co-current beam correlated strongly on the local electron density and related parameters (plasma current and neutron rate) and increased with deeper fast-ion loss boundaries towards the magnetic axis. This was also the case with second harmonic ICE driven by the counter-current beam but not with its stronger third harmonic emission. The central ICE harmonics destabilized by both beams are observed to have different temporal dynamics. The central ICE amplitude responds rapidly to transient MHD events; it dropped and recovered in less than a millisecond at each sawtooth event. ICE frequency splitting is triggered by both the co-current and counter-current 'near-tangential' beams. The data presented provide opportunities to test and validate models of excitation of ICE by energetic ions.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

El-CID: a filter for gravitational-wave electromagnetic counterpart identification

ABSTRACT As gravitational-wave (GW) interferometers become more sensitive and probe ever more distant reaches, the number of detected binary neutron star mergers will increase. However, detecting more events farther away with GWs does not guarantee corresponding increase in the number of electromagnetic counterparts of these events. Current and upcoming wide-field surveys that participate in GW follow-up operations will have to contend with distinguishing the kilonova (KN) from the ever increasing number of transients they detect, many of which will be consistent with the GW sky-localization. We have developed a novel tool based on a temporal convolutional neural network architecture, trained on sparse early-time photometry and contextual information for Electromagnetic Counterpart Identification (El-CID). The overarching goal for El-CID is to slice through list of new transient candidates that are consistent with the GW sky localization, and determine which sources are consistent with KNe, allowing limited target-of-opportunity resources to be used judiciously. In addition to verifying the performance of our algorithm on an extensive testing sample, we validate it on AT2017gfo – the only EM counterpart of a binary neutron star merger discovered to date – and AT2019npv – a supernova that was initially suspected as a counterpart of the GW event, GW190814, but was later ruled out after further analysis.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2023

ML-Based Power System Stability Assessment Considering Network Topology Changes: WECC 20,000+ Bus System Case Study

Modern power grids are fast-changing and thus require real-time monitoring and online stability assessment. With the rapid development of machine learning (ML) techniques, using data-driven models to provide fast and accurate estimations of power system stability marginal information, such as frequency nadir for frequency stability and critical clearing time (CCT) for transient stability, have become possible. However, despite the numerous research on ML-based methods for frequency nadir and CCT prediction, there is limited work on the impact of different network topology changes. Furthermore, most previous studies only focused on small or synthetic systems, and there is a lack of research on actual large power system models. In this paper, the above issues are addressed by studying the actual U.S. Western Electricity Coordinating Council (WECC) system model with more than 20,000 buses. Massive simulations are conducted in PowerWorld Simulator to study the impact of various topology change scenarios on both frequency stability and transient stability. System operating information is extracted from the success dispatch cases of various network topologies to generate a comprehensive dataset for ML-based models. Two ML methods, random forest (RF) and multilayer perceptron (MLP) neural network, are trained and tested for both frequency nadir prediction and CCT prediction. Test results have proven the models are capable of online stability assessment for large power networks such as the WECC system with sufficient accuracy.

critical clearing time↗
DOE OSTI2012

Blister Threshold Based Thermal Limits for the U-Mo Monolithic Fuel System

Fuel failure is most commonly induced in research and test reactor fuel elements by exposure to an under-cooled or over-power condition that results in the fuel temperature exceeding a critical threshold above which blisters form on the plate. These conditions can be triggered by normal operational transients (i.e. temperature overshoots that may occur during reactor startup or power shifts) or mild upset events (e.g., pump coastdown, small blockages, mis-loading of fuel elements into higher-than-planned power positions, etc.). The rise in temperature has a number of general impacts on the state of a fuel plate that include, for example, stress relaxation in the cladding (due to differential thermal expansion), softening of the cladding, increased mobility of fission gases, and increased fission-gas pressure in pores, all of which can encourage the formation of blisters on the fuel-plate surface. These blisters consist of raised regions on the surface of fuel plates that occur when the cladding plastically deforms in response to fission-gas pressure in large pores in the fuel meat and/or mechanical buckling of the cladding over damaged regions in the fuel meat. The blister temperature threshold decreases with irradiation because the mechanical properties of the fuel plate degrade while under irradiation (due to irradiation damage and fission-product accumulation) and because the fission-gas inventory progressively increases (and, thus, so does the gas pressure in pores).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

A Robust Yule-Walker Method for Online Monitoring of Power System Electromechanical Modes of Oscillation

Reliable power system operation requires that system operators maintain a satisfactory small-signal stability margin at all times. Mode meters allow system operators to continuously monitor the stability margin by analyzing synchrophasor measurements. Experience with mode meters at the Bonneville Power Administration (BPA) has revealed that mode meters sometimes suffer severe bias following transient disturbances. In this paper, a robust version of the Yule-Walker mode meter algorithm is proposed to mitigate this bias. The robust capability is introduced by modifying the estimator of the autocovariance sequence (ACS) on which the algorithm depends. Tests with field-measured data demonstrate that the robust Yule-Walker (RYW) algorithm significantly outperforms its conventional counterpart and is better able to provide system operators with accurate information about the system's stability margin.

Follum, James D.↗
DOE OSTI2016

A Lyapunov Function Based Remedial Action Screening Tool Using Real-Time Data

This report summarizes the outcome of a research project that comprised the development of a Lyapunov function based remedial action screening tool using real-time data (L-RAS). The L-RAS is an advanced computational tool that is intended to assist system operators in making real-time redispatch decisions to preserve power grid stability. The tool relies on screening contingencies using a homotopy method based on Lyapunov functions to avoid, to the extent possible, the use of time domain simulations. This enables transient stability evaluation at real-time speed without the use of massively parallel computational resources. The project combined the following components. 1. Development of a methodology for contingency screening using a homotopy method based on Lyapunov functions and real-time data. 2. Development of a methodology for recommending remedial actions based on the screening results. 3. Development of a visualization and operator interaction interface. 4. Testing of screening tool, validation of control actions, and demonstration of project outcomes on a representative real system simulated on a Real-Time Digital Simulator (RTDS) cluster. The project was led by Michigan State University (MSU), where the theoretical models including homotopy-based screening, trajectory correction using real-time data, and remedial action were developed and implemented in the form of research-grade software. Los Alamos National Laboratory (LANL) contributed to the development of energy margin sensitivity dynamics, which constituted a part of the remedial action portfolio. Florida State University (FSU) and Southern California Edison (SCE) developed a model of the SCE system that was implemented on FSU's RTDS cluster to simulate real-time data that was streamed over the internet to MSU where the L-RAS tool was executed and remedial actions were communicated back to FSU to execute stabilizing controls on the simulated system. LCG Consulting developed the visualization and operator interaction interface, based on specifications provided by MSU. The project was performed from October 2012 to December 2016, at the end of which the L-RAS tool, as described above, was completed and demonstrated. The project resulted in the following innovations and contributions: (a) the L-RAS software prototype, tested on a simulated system, vetted by utility personnel, and potentially ready for wider testing and commercialization; (b) an RTDS-based test bed that can be used for future research in the field; (c) a suite of breakthrough theoretical contributions to the field of power system stability and control; and (d) a new tool for visualization of power system stability margins. While detailed descriptions of the development and implementation of the various project components have been provided in the quarterly reports, this final report provides an overview of the complete project, and is demonstrated using public domain test systems commonly used in the literature. The SCE system, and demonstrations thereon, are not included in this report due to Critical Energy Infrastructure Information (CEII) restrictions.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2019

Results of the Phenix Dissymmetric Test Benchmark Exercise

In the context of the European H2020 project SESAME, a work package was dedicated to the validation of code coupling for the thermal-hydraulics of pool type reactors. To this end, CEA proposed a benchmark exercise of the so-called dissymmetric test performed at the end of life of the PHENIX reactor. The PHENIX reactor was an experimental reactor operated by CEA in France from 1974 until 2009. The knowledge and data acquired over the lifetime of this reactor are of paramount importance for the development of future sodium-cooled generation IV reactors. At the beginning of this test, the reactor is operated at nominal power (350 MWth) with 3 primary pumps and 4 intermediate heat exchangers. During this test, a pump trip occurs at the secondary side of one of the two loops (one loop provides coolant to two intermediate heat exchangers). The chain reaction is then stopped while the pumps remain operating at nominal speed. Dissymmetric flow patterns are then observed in the cold pool as hot sodium leaves the intermediate heat exchangers that are no longer cooled. The failed secondary pump is then restarted while the rotational speed of the other is reduced. Eventually, a stable homogeneous state is reached. This transient is challenging for both modelers and codes since this exercise tackles an industrial scale reactor featuring complex flows and interactions between different scales. The physical phenomena observed include thermal stratification, flow mixing and buoyancy effects. In order to accurately model these phenomena for a reasonable computational costs, the partners used System Thermal Hydraulics codes coupled with Computational Fluid Dynamics codes and one even included a subchannel code. In this paper, the PHENIX reactor and its behavior during the dissymmetric test are described. Then, a comparison of the different modeling options selected by the partners is offered. Finally, the results are compared to the experimental data and discussed and modelling recommendations are proposed.

code coupling↗
DOE OSTI2022

Representative-cell-based Modeling of HTTF in SAM

Although the current nuclear power market is primarily occupied by the light water reactors (LWRs), the TRISO-fueled helium-cooled graphite-moderated high temperature gas reactors (HTGRs) are drawing growing attention as the nuclear power industry marches towards more advanced systems. Compared to that of conventional LWRs, the thermal hydraulics of HTGR cores show extra complexity from their multi-scale heat transfer mechanisms with varying importance during different operation modes or transient stages. The modeling of HTGR cores for the system analysis purpose therefore faces the challenge of reaching reasonable fidelity and accuracy while maintaining sufficient simplicity. Previous development and validation efforts have demonstrated that a “2-D ring model” with reasonable performance can be implemented in the System Analysis Module (SAM) for prismatic HTGR cores. In the current project, an alternative modeling methodology based on “representative cells” is proposed for the typical prismatic core of HTGRs. Different from the previous ring model approach, the proposed methodology first separately models and then combines the small- and large-scale thermal conductions, by connecting representative local heat transfer units (cells) with effective core- wise thermal resistance. A modeling practice for an integral high-temperature test facility (HTTF) elaborates the modeling details. Steady-state validation of the resultant candidate model is performed against a higher-resolution benchmark from a 3D-1D coupled simulation, which shows satisfactory prediction with reasonably captured global parameters and well-represented temperature fields. A postulated pressurized conduction cooldown (PCC) is also simulated and analyzed, demonstrating the model’s capability of transient prediction with physically captured phenomena resolved in both small and large temporal and spatial scales. In general, the current work achieves a preliminary success in proposing a methodology using representative cells to model prismatic HTGR cores in SAM. Future efforts are envisioned with more validation activities and with potential modeling improvements to eventually achieve the high confidence on a high-fidelity robust modeling methodology with reasonable accuracy.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Validation of MFUEL Metal Fuel Performance Models of SAS4A/SASSYS-1

The fuel characterization models of SAS4A/SASSYS-1 (SAS) have recently been extended to include a new U-Pu-Zr metal fuel model, MFUEL. MFUEL is equipped with mechanistic physics based models to predict the pre-transient characterization and transient response of metal fuel, with emphasis on fuel melting, cladding failure, and the metal fuel’s impact on core reactivity. The MFUEL model will be available in the full version of SAS4A/SASSYS-1 5.7, which is scheduled to be released in June 2023. Fast reactor fuel pins that operated in EBR-II and FFTF with low smear density U-Zr and U-Pu-Zr metal fuels and irradiation resistant ferritic-martensitic cladding showed significant advantages in achieving high burnups and assuring inherent safety characteristics during anticipated transients, design basis events, and beyond design basis events. For safety analysis, a fuel performance model must be able to predict (1) Fuel pin mechanics and compositional and dimensional changes, (2) Clad failure, and (3) Fuel pin thermal resistance. Achieving these high level goals accurately is strongly related to the model performance of individual physical processes taking place within a fuel pin during its lifetime. Metal fuels typically operate above the mid-point temperature of melting during normal, as well as off-normal, conditions. At these elevated temperatures, the availability of thermal activation provides a driving force for various diffusional processes leading to complex phase transformations, micro-structure evolution, significant amounts of fuel swelling, interconnected porosity formation, excessive amounts of fission gas release, and fuel clad chemical interactions. Clad failure in fast reactors primarily occurs as a result of creep rupture augmented by clad wastage formation. The reaction is driven by thermal creep induced dislocation motion, grain boundary cavity nucleation, growth and breakup of grain boundaries. The high level complexity and limited available data requires introducing physics-based modeling approaches to gain extrapolation ability and sensitivity with respect to various conditions. The objective of this report is to perform validation of MFUEL using the experimental data for (1) Normal operation EBR-II fuel behavior, (2) Normal operation PHENIX fuel behavior, (3) HT9 Pressure tube ramp-and-hold creep rupture tests, (4) Whole Pin Furnace (WPF) creep strain, creep rupture and eutectic tests, (5) Fuel Behavior Test Apparatus (FBTA) eutectic tests, and (6) TREAT M5-7 OverPower tests up to clad failure. Section-2 includes a brief description of the MFUEL models. A detailed description of the MFUEL physics-based, semi-empirical models will be presented in the SAS V 5.7 theory manual. Section-3, 4, and 5 describes the validation effort for the pre-transient irradiation, furnace transients, and TREAT M-Series transients, respectively.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Progress on Computational Modeling of Water-Based NSTF (FY2021)

This report documents the FY21 progress and achievements made in the computational analyses of the water-based NSTF. Both system-level and high-fidelity Computational Fluid Dynamics (CFD) analyses were performed to gain a complete understanding of the complex flow and heat transfer phenomena in natural convection systems. The progress on the waterbased NSTF experimental testing is summarized in a companion report (ANL-ART-230). As a continuation of progress from previous years, in FY21 the RELAP5 model of the NSTF was first updated to include a representative heated cavity that incorporated convection and conduction means of heat transfer. Two different cavity models were developed, one that models the radiation heat exchange among all surfaces in one enclosure (Cavity Model 1), and another that employs multiple radiation enclosures with each considering the surfaces at one axial level (Cavity Model 2). This updated RELAP5 deck was benchmarked with single-phase test data to tune form loss coefficients of the elbows and tees, along with heat transfer coefficients within the cavity and off the external insulation panels. This tuned model was then found to accurately predict fluid temperatures within 2.5% and both system and riser flow rates within 6% during single-phase, steady-state simulations. For transient simulations, there is a slight over prediction of the flow rate, fluid temperatures, and heater temperatures during the heat-up period, which indicate that the transient heat loss may be underestimated. Following, the capabilities of RELAP5-3D were examined, which feature a conduction enclosure model that is not present in the current version of RELAP5-MOD3.3 used for all previous NSTF simulations. As an initial comparison of the two codes, single-phase simulations were run that used similar input decks. The results indicate that the single-phase flow predicted by the two codes was very similar, however RELAP5-3D appears to predict a higher system flow during the two-phase transient when compared to the flow rates predicted by RELAP5-MOD3.3. Furthermore, RELAP5-3D run did not capture the density wave oscillations or geysering phenomena, which are large oscillation flow instabilities observed in the experimental facility. To further the prediction capabilities for capturing two-phase instabilities, Options 55, 57, 58, 61, which are implemented by default in RELAP5-MOD3.3 but not in RELAP5-3D, were investigated. When enabling Option 61, RELAP5-3D agrees well with MOD3.3 in predicting the total system flow, while other options have negligible effect on the predicted total system flow by RELAP5-3D. However, none of these options enables RELAP5-3D to produce the same flow instabilities. For future steps, efforts are currently underway to investigate RELAP5-3D’s different behavior in predicting low-pressure two-phase instabilities. STAR-CCM+ was used to perform CFD analysis of the water-based test facility and provide simulation results for comparisons to experimental data. Efforts first focused on describing the characteristics of flow within the heated cavity. A study was performed on whether possible leakage flow between the two sides of the cavity could substantially impact and explain the results. An additional second assumption was examined which defined adiabatic side walls, so physical insulation structures were added to the models to allow conduction across this outer insulation. The results suggest that potential porosity of the insulation could be a major factor in the temperature distribution within the cavity and should be considered in any studies where it is desirable to know the cavity wall temperatures to higher accuracy.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗