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At least 1,063 records · Page 59

The international implications of national and local coordination on building energy codes: Case studies in six cities

Building energy efficiency is an important strategy for reducing greenhouse gas emissions globally. In fact, 55 countries have included building energy efficiency in their Nationally Determined Contributions (NDCs) under the Paris Agreement. This research uses building energy code implementation in six cities across different continents as case studies to assess what it may take for countries to implement the ambitions of their energy efficiency goals. Specifically, we look at the cases of Bogota, Colombia; Da Nang, Vietnam; Eskisehir, Turkey; Mexico City, Mexico; Rajkot, India; and Tshwane, South Africa, all of which are “deep dive” cities under the Sustainable Energy for All's Building Efficiency Accelerator. The research focuses on understanding the baseline with existing gaps in implementation and coordination. The methodology used a combination of surveys on code status and interviews with stakeholders at the local and national level, as well as review of published documents. We looked at code development, implementation, and evaluation. The cities are all working to improve implementation, however, the challenges they currently face include gaps in resources, capacity, tools, and institutions to check for compliance. Better coordination between national and local governments could help improve implementation, but that coordination is not yet well established. For example, all six of the cities reported that there was little to no involvement of local stakeholders in development of the national code; only one city reported that it had access to national funding to support code implementation. More robust coordination could better link cities with capacity building and funding for compliance, and ensure that the code reflects local priorities. By understanding gaps in implementation, it can also help in designing more targeted interventions to scale up energy savings.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

Characteristics of Vertical Ground Motions and Their Effect on the Seismic Response of Bridges in the Near-Field: A State-of-the-Art Review

Despite the evidence from past earthquakes and several numerical investigations demonstrating the detrimental impact of vertical ground motions (VGMs) on the integrity of bridge structures, incorporating their effects into seismic assessment and design procedures has traditionally been given limited consideration. Current codes utilize rather simplistic approaches to account for the concurrent effects of vertical and horizontal motions in structural performance evaluations, potentially leading to unconservative estimates of structural demands. This paper reviews the main features of VGMs and their effect on the seismic response of bridges. The methods and empirical models available to estimate vertical motions for design purposes are discussed, and research gaps and related research needs are identified. Finally, the emerging role of physics-based ground-motion simulations, as well as their limitations, in supporting future research and informing the development of simplified design procedures is examined. The main areas of interest for future research are identified in the need to carry out systematic sensitivity studies to gain insight into the main earthquake parameters that influence key VGMs features, understand the influence of soil nonlinearities on VGMs amplitude and frequency content, inform the development of empirical models that cover a range of site conditions and source-to-site distances where current models are poorly constrained, generate arrays of motions to update coherency models to properly inform the analysis of distributed infrastructure, investigate the impulsive character of VGMs, and assess the approximations made in estimating VGMs with 1D site response analyses. Furthermore, specific focus is laid on large-magnitude earthquakes in the near-field.

Asynchronous ground motion↗

The ITER 3D Magnetic Diagnostic Response to Applied n=3 and n=4 RMP's

The ITER magnetic diagnostic response to applied n=3 and n=4 RMPs has been calculated for the 15MA scenario. The VMEC code was utilized to calculate free boundary 3D ideal MHD equilibria, where the non-stellarator symmetric terms were included in the calculation. This allows an assessment to be made of the possible boundary displacements due to RMP application in ITER. As the VMEC code assumes a continuous set of nested flux surface, the possibility of island and stochastic region formation is ignored. At the start of the current at-top (L-Mode) application of n = 4 RMP's indicates approximately 1 cm peak-to-peak displacements on the low field side of the plasma while later in the shot (H-mode) perturbations as large as 3 cm are present. Forward modeling of the ITER magnetic diagnostics indicates significant non-axisymmetric plasma response, exceeding 10% the axisymmetric signal in many of the flux loops. Magnetic field probes seem to indicate a greater robustness to 3D effects but still indicate large sensitivities to 3D effects in a number of sensors. Forward modeling of the diagnostics response to 3D equilibria allows assessment of diagnostics design and control scenarios.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Fracture Capabilities in Grizzly with the extended Finite Element Method (X-FEM)

Efforts are underway to develop fracture mechanics capabilities in the Grizzly code to enable it to be used to perform deterministic fracture assessments of degraded reactor pressure vessels (RPVs). A capability was previously developed to calculate three-dimensional interaction- integrals to extract mixed-mode stress-intensity factors. This capability requires the use of a finite element mesh that conforms to the crack geometry. The eXtended Finite Element Method (X-FEM) provides a means to represent a crack geometry without explicitly fitting the finite element mesh to it. This is effected by enhancing the element kinematics to represent jump discontinuities at arbitrary locations inside of the element, as well as the incorporation of asymptotic near-tip fields to better capture crack singularities. In this work, use of only the discontinuous enrichment functions was examined to see how accurate stress intensity factors could still be calculated. This report documents the following work to enhance Grizzly’s engineering fracture capabilities by introducing arbitrary jump discontinuities for prescribed crack geometries; X-FEM Mesh Cutting in 3D: to enhance the kinematics of elements that are intersected by arbitrary crack geometries, a mesh cutting algorithm was implemented in Grizzly. The algorithm introduces new virtual nodes and creates partial elements, and then creates a new mesh connectivity; Interaction Integral Modifications: the existing code for evaluating the interaction integral in Grizzly was based on the assumption of a mesh that was fitted to the crack geometry. Modifications were made to allow for the possibility of a crack front that passes arbitrarily through the mesh; and Benchmarking for 3D Fracture: the new capabilities were benchmarked against mixed-mode three-dimensional fracture problems with known analytical solutions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

RAVEN User Manual

RAVEN is a generic software framework to perform parametric and probabilistic analysis based on the response of complex system codes. The initial development was aimed to provide dynamic risk analysis capabilities to the Thermo-Hydraulic code RELAP-7, currently under development at the Idaho National Laboratory (INL). Although the initial goal has been fully accomplished, RAVEN is now a multi-purpose probabilistic and uncertainty quantification platform, capable to agnostically communicate with any system code. This agnosticism includes providing Application Programming Interfaces (APIs). These APIs are used to allow RAVEN to interact with any code as long as all the parameters that need to be perturbed are accessible by input files or via python interfaces. RAVEN is capable of investigating the system response, and investigating the input space using Monte Carlo, Grid, or Latin Hyper Cube sampling schemes, but its strength is focused toward system feature discovery, such as limit surfaces, separating regions of the input space leading to system failure, using dynamic supervised learning techniques. The development of RAVEN started in 2012, when, within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the need to provide a modern risk evaluation framework became stronger. RAVEN principal assignment is to provide the necessary software and algorithms in order to employ the concept developed by the Risk Informed Safety Margin Characterization (RISMC) program. RISMC is one of the pathways defined within the Light Water Reactor Sustainability (LWRS) program. In the RISMC approach, the goal is not just the individuation of the frequency of an event potentially leading to a system failure, but the closeness (or not) to key safety-related events. Hence, the approach is interested in identifying and increasing the safety margins related to those events. A safety margin is a numerical value quantifying the probability that a safety metric (e.g. for an important process such as peak pressure in a pipe) is exceeded under certain conditions. The initial development of RAVEN has been focused on providing dynamic risk assessment capability to RELAP-7, currently under development at the INL and, likely, future replacement of the RELAP5-3D code. Most the capabilities that have been implemented having RELAP-7 as principal focus are easily deployable for other system codes. For this reason, several side activates are currently ongoing for coupling RAVEN with software such as RELAP5-3D, etc. The aim of this document is the explanation of the input requirements, focusing on the input structure.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Qualitative Risk Assessment of Legacy Wells within the Estimated Prairie State Generating Company Area of Review

This report details the digitization of a legacy wellbore database, including data processing assumptions, parameter estimation, and risk assessment methodology. The database, comprising 6,454 documents, was provided by ISGS. It includes valuable data from the Prairie State Generating Company (PSGC) and One Earth Energy (OEE) sites of the CarbonSAFE Phase III – Illinois Storage Corridor project. The report focuses on wells within a 15-mile radius from the Lively Grove #1 (LG#1) well at PSGC site, evaluating subsurface conditions and potential risks. A total of 4,386 wellbores within 15 miles of the LG#1 well were filtered based on depth and formation codes. LG#1 is the stratigraphic well at the PSGC site drilled in 2021. Ninety-four (94) wells penetrating the Maquoketa Shale Group (the primary confining unit) within the estimated area-of-review (AoR) for the PSGC site were evaluated using a qualitative risk assessment (QRA) methodology. The QRA developed by Arbad et al. 2022 focuses on legacy wells within the AoR and categorizes them based on well construction details. The QRA identifies wells that need immediate attention by categorizing them based on penetration depth and protection. Wells within the AoR were categorized into nine groups based on penetrations and protections. These categories range from Type 1 wells, with no documentation, to Type 9 wells, which do not penetrate the primary confining unit or storage reservoir (unit). Well accessibility within the AoR varies based on well status, including Dry & Abandoned (DA), Plugged & Abandoned (PA), Injection (INJ), Oil/Gas Producing (PROD), and Observation (Obs) wells. Accessibility levels were determined by well construction, with DA wells being the least accessible and Observation wells the most accessible, impacting gas leakage detection possibilities. Remedial action priority of wells decreases from Type 1 to Type 9 wells. Type 1 to Type 6 wells with status DA and PA require immediate attention, while Type 7 and Type 8 wells are low priority. A risk matrix used to prioritize corrective actions for legacy wells is proposed to categorize wells within an AoR based on penetrations, protections, and accessibility. The methodology involves data acquisition, well categorization into nine types, and determining CO 2 leakage pathways using well schematics and geospatial mapping. This approach is particularly useful for managing the integrity of legacy wells throughout the lifecycle of a Carbon Capture and Storage (CCS) project. A qualitative risk assessment of 94 wells within the AoR of the PSGC site identified 54 wells with high priority for corrective action due to penetration of the primary containment seal. The assessment utilizes color-coded maps to categorize well types and prioritize corrective actions, providing a comprehensive analysis. Schematics of wells penetrating the primary confining unit were drawn, and leakage pathways were identified. Details of all wells penetrating the confining zone are provided in the appendix, including information on well types, plugging, and casing status.

01 COAL, LIGNITE, AND PEAT↗

Reactor Pressure Vessel Fracture Analysis Capabilities in Grizzly

Efforts have been underway to develop fracture mechanics capabilities in the Grizzly code to enable it to be used to perform deterministic fracture assessments of degraded reactor pressure vessels (RPVs). Development in prior years has resulted a capability to calculate -integrals. For this application, these are used to calculate stress intensity factors for cracks to be used in deterministic linear elastic fracture mechanics (LEFM) assessments of fracture in degraded RPVs. The -integral can only be used to evaluate stress intensity factors for axis-aligned flaws because it can only be used to obtain the stress intensity factor for pure Mode I loading. Off-axis flaws will be subjected to mixed-mode loading. For this reason, work has continued to expand the set of fracture mechanics capabilities to permit it to evaluate off-axis flaws. This report documents the following work to enhance Grizzly’s engineering fracture mechanics capabilities for RPVs: • Interaction Integral and -stress: To obtain mixed-mode stress intensity factors, a capability to evaluate interaction integrals for 2D or 3D flaws has been developed. A -stress evaluation capability has been developed to evaluate the constraint at crack tips in 2D or 3D. Initial verification testing of these capabilities is documented here. • Benchmarking for axis-aligned flaws: Grizzly’s capabilities to evaluate stress intensity factors for axis-aligned flaws have been benchmarked against calculations for the same conditions in FAVOR. • Off-axis flaw demonstration: The newly-developed interaction integral capabilities are demon- strated in an application to calculate the mixed-mode stress intensity factors for off-axis flaws. • Other code enhancements: Other enhancements to the thermomechanics capabilities that relate to the solution of the engineering RPV fracture problem are documented here.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

An Assessment of MELCOR 2.1: Containment Thermal-Hydraulic Tests in the Heissdampfreaktor (HDR) Facility

MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. It provides a capability for independently auditing analyses submitted by reactor manufacturers and utilities. In order to assess the adequacy of containment thermal - hydraulic modeling incorporated in the MELCOR code, a key containment test facility was analyzed. This report documents MELCOR code calculations for simulating steam - water blowdown tests performed in the Heissdampfreaktor ( HDR) de-commissioned containment facility located near Frankfurt , Germany . These tests are a series of blowdown experiments in a large scaled test facility ; including some tests with the addition of hydrogen release which are intended to simulate a variety of postulated break s inside large containment buildings. The key objectives of this MELCOR assessment are to study: (1) the expansion and transport of high energy steam - water releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment gas mixing and stratification. Moreover, MELCOR results are compared to the CONTAIN code for the same tests.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

A Computational Study of an Oscillating VR-12 Airfoil with a Gurney Flap

Computations of the flow over an oscillating airfoil with a Gurney-flap are performed using a Reynolds Averaged Navier-Stokes code and compared with recent experimental data. The experimental results have been generated for different sizes of the Gurney flaps. The computations are focused mainly on a configuration. The baseline airfoil without a Gurney flap is computed and compared with the experiments in both steady and unsteady cases for the purpose of initial testing of the code performance. The are carried out with different turbulence models. Effects of the grid refinement are also examined and unsteady cases, in addition to the assessment of solver effects. The results of the comparisons of steady lift and drag computations indicate that the code is reasonably accurate in the attached flow of the steady condition but largely overpredicts the lift and underpredicts the drag in the higher angle steady flow.

Rhee, Myung↗

Proceedings of the Third International Workshop on Proof-Carrying Code and Software Certification

This NASA conference publication contains the proceedings of the Third International Workshop on Proof-Carrying Code and Software Certification, held as part of LICS in Los Angeles, CA, USA, on August 15, 2009. Software certification demonstrates the reliability, safety, or security of software systems in such a way that it can be checked by an independent authority with minimal trust in the techniques and tools used in the certification process itself. It can build on existing validation and verification (V&V) techniques but introduces the notion of explicit software certificates, Vvilich contain all the information necessary for an independent assessment of the demonstrated properties. One such example is proof-carrying code (PCC) which is an important and distinctive approach to enhancing trust in programs. It provides a practical framework for independent assurance of program behavior; especially where source code is not available, or the code author and user are unknown to each other. The workshop wiII address theoretical foundations of logic-based software certification as well as practical examples and work on alternative application domains. Here "certificate" is construed broadly, to include not just mathematical derivations and proofs but also safety and assurance cases, or any fonnal evidence that supports the semantic analysis of programs: that is, evidence about an intrinsic property of code and its behaviour that can be independently checked by any user, intermediary, or third party. These guarantees mean that software certificates raise trust in the code itself, distinct from and complementary to any existing trust in the creator of the code, the process used to produce it, or its distributor. In addition to the contributed talks, the workshop featured two invited talks, by Kelly Hayhurst and Andrew Appel. The PCC 2009 website can be found at http://ti.arc.nasa.gov /event/pcc 091.

Ewen, Denney, W.↗

Assessment of Macroscopic k-e Model of CUPID for a Subchannel Analysis

Subchannel analysis codes have been widely used for the rod bundle analysis in reactor core. In the subchannel analysis, turbulent mixing which account for cross flows is very important and has to be modeled considering different flow geometry. In this study, macroscopic k-e turbulence model is used for the replacement of the turbulence mixing term. However, since the macroscopic k-e turbulence model is developed mainly for sphere-packed bed and channel or pipe bundle problems, for the flows including strong cross flows which contain strong dispersion effect the macroscopic models should be validated. This paper aims to explain the validity of the macroscopic k-e turbulence model of the CUPID code with two types of rod bundle validation problems. In this study, the macroscopic k-e model has been validated for the application to subchannel analysis. It has been implemented in the CUPID code and validated against CNEN 4x4 and PNL 7x7 rod bundle tests. The results showed that the macroscopic k-e turbulence model can estimate the experiments properly. Especially when the major flow direction is longitudinal to the rod bundle, the macroscopic turbulence model was the same with the subchannel turbulence mixing model. For the flows across the blockage, the strong dispersion around the blockage can may affect the deviation of velocity. For the future work, the macroscopic k-e turbulence model will be modified to include the dispersion effect. (authors)

42 ENGINEERING↗

Development and Testing of Neutron Cross Section Covariance Data for SCALE 6.2

Neutron cross-section covariance data are essential for many sensitivity/uncertainty and uncertainty quantification assessments performed both within the TSUNAMI suite and more broadly throughout the SCALE code system. The release of ENDF/B-VII.1 included a more complete set of neutron cross-section covariance data: these data form the basis for a new cross-section covariance library to be released in SCALE 6.2. A range of testing is conducted to investigate the properties of these covariance data and ensure that the data are reasonable. These tests include examination of the uncertainty in critical experiment benchmark model k eff values due to nuclear data uncertainties, as well as similarity assessments of irradiated pressurized water reactor (PWR) and boiling water reactor (BWR) fuel with suites of critical experiments. The contents of the new covariance library, the testing performed, and the behavior of the new covariance data are described in this paper. The neutron cross-section covariances can be combined with a sensitivity data file generated using the TSUNAMI suite of codes within SCALE to determine the uncertainty in system k eff caused by nuclear data uncertainties. The Verified, Archived Library of Inputs and Data (VALID) maintained at Oak Ridge National Laboratory (ORNL) contains over 400 critical experiment benchmark models, and sensitivity data are generated for each of these models. The nuclear data uncertainty in k eff is generated for each experiment, and the resulting uncertainties are tabulated and compared to the differences in measured and calculated results. The magnitude of the uncertainty for categories of nuclides (such as actinides, fission products, and structural materials) is calculated for irradiated PWR and BWR fuel to quantify the effect of covariance library changes between the SCALE 6.1 and 6.2 libraries. One of the primary applications of sensitivity/uncertainty methods within SCALE is the assessment of similarities between benchmark experiments and safety applications. This is described by a c k value for each experiment with each application. Several studies have analyzed typical c k values for a range of critical experiments compared with hypothetical irradiated fuel applications. The c k value is sensitive to the cross-section covariance data because the contribution of each nuclide is influenced by its uncertainty; large uncertainties indicate more likely bias sources and are thus given more weight. Changes in c k values resulting from different covariance data can be used to examine and assess underlying data changes. These comparisons are performed for PWR and BWR fuel in storage and transportation systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Development and testing of neutron cross-section covariance data for Scale 6.2 - 14604

Neutron cross-section covariance data are essential for many sensitivity/uncertainty and uncertainty quantification assessments performed both within the TSUNAMI suite and more broadly throughout the SCALE code system. The release of ENDF/B-VII.1 included a more complete set of neutron cross-section covariance data: these data form the basis for a new cross-section covariance library to be released in SCALE 6.2. A range of testing is conducted to investigate the properties of these covariance data and ensure that the data are reasonable. These tests include examination of the uncertainty in critical experiment benchmark model k{sub eff} values due to nuclear data uncertainties, as well as similarity assessments of irradiated pressurized water reactor (PWR) and boiling water reactor (BWR) fuel with suites of critical experiments. The contents of the new covariance library, the testing performed, and the behavior of the new covariance data are described in this paper. The neutron cross-section covariances can be combined with a sensitivity data file generated using the TSUNAMI suite of codes within SCALE to determine the uncertainty in system k{sub eff} caused by nuclear data uncertainties. The Verified, Archived Library of Inputs and Data (VALID) maintained at Oak Ridge National Laboratory (ORNL) contains over 400 critical experiment benchmark models, and sensitivity data are generated for each of these models. The nuclear data uncertainty in k{sub eff} is generated for each experiment, and the resulting uncertainties are tabulated and compared to the differences in measured and calculated results. The magnitude of the uncertainty for categories of nuclides (such as actinides, fission products, and structural materials) is calculated for irradiated PWR and BWR fuel to quantify the effect of covariance library changes between the SCALE 6.1 and 6.2 libraries. One of the primary applications of sensitivity/uncertainty methods within SCALE is the assessment of similarities between benchmark experiments and safety applications. This is described by a ck value for each experiment with each application. Several studies have analyzed typical ck values for a range of critical experiments compared with hypothetical irradiated fuel applications. The ck value is sensitive to the cross-section covariance data because the contribution of each nuclide is influenced by its uncertainty; large uncertainties indicate more likely bias sources and are thus given more weight. Changes in ck values resulting from different covariance data can be used to examine and assess underlying data changes. These comparisons are performed for PWR and BWR fuel in storage and transportation systems. (authors)

42 ENGINEERING↗

ET TPS Development Criteria and Flight Instrumentation Program

The objective was to verify the design thermal environments for the Space Shuttle External Tank. These environments encompassed a myraid of potential flight cases which included nominal and dispersed trajectories launched from Kennedy Space Center (KSC) and Vandenberg and abort cases. Many iterations of environments have been produced during the evolution of the Space Shuttle system which required an analysis and assessment of each environment. A significant effort was expended in the analysis and application of wind tunnel and heat transfer data to develop math/computer models. Various versions of the aeroheating computer code MINIVER were tailored to calculate ascent aeroheating environments for the External Tank. The significant environment assessment results are summarized.

Warmbrod, John D.↗

Studying visual search without an eye tracker: an assessment of artificial foveation

Abstract Eye tracking is a useful tool for studying human cognition, both in the laboratory and in real-world applications. However, there are cases in which eye tracking is not possible, such as in high-security environments where recording devices cannot be introduced. After facing this challenge in our own work, we sought to test the effectiveness of using artificial foveation as an alternative to eye tracking for studying visual search performance. Two groups of participants completed the same list comparison task, which was a computer-based task designed to mimic an inventory verification process that is commonly performed by international nuclear safeguards inspectors. We manipulated the way in which the items on the inventory list were ordered and color coded. For the eye tracking group, an eye tracker was used to assess the order in which participants viewed the items and the number of fixations per trial in each list condition. For the artificial foveation group, the items were covered with a blurry mask except when participants moused over them. We tracked the order in which participants viewed the items by moving their mouse and the number of items viewed per trial in each list condition. We observed the same overall pattern of performance for the various list display conditions, regardless of the method. However, participants were much slower to complete the task when using artificial foveation and had more variability in their accuracy. Our results indicate that the artificial foveation method can reveal the same pattern of differences across conditions as eye tracking, but it can also impact participants’ task performance.

47 OTHER INSTRUMENTATION↗

Probabilistic Seismic Hazard Assessment for Georgia

Probabilistic Seismic Hazard Assessments (PSHA) form the basis for most contemporary seismic provisions in building codes around the world. The current building code of Georgia (2009) is not fully compatible with seismic hazard and design concepts of modern codes. The Ministry of Economy and Sustainable Development of Georgia is currently undertaking an update to the building code to align it with EuroCode. In conjunction with EuroCode, a National Annex specific to Georgia is needed, including seismic hazard values for the territory of Georgia. A comprehensive earthquake catalogue is fundamental to a robust PSHA in the region. The Caucasus has had an uneven seismic recording and reporting history both temporally and spatially. Therefore, there has been a strong regional effort to bring together data from all the countries in the Caucasus, to digitize and standardize paper-based data from the Soviet era, and to re-examine magnitude and location calculations. The earthquake catalogue resulting from this effort forms the basis for a reliable seismic hazard assessment. Recognizing these needs, the U.S. Department of Energy through Lawrence Livermore National Laboratory (LLNL) initiated a project in 2015 to engage and train local scientists in the Caucasus to install new equipment, to enhance the quality of seismic monitoring and reporting, and to improve and harmonize the regional earthquake catalogue. In 2016, the project scope was expanded to include the modernization of seismic hazard assessments. To this end, a series of workshops were held in Georgia, the first one of which presented an overview of PSHA and its components to various participants from the universities and other research organizations in the region. This kick-off workshop was followed by numerous hands-on training workshops in 2017 and 2018, covering all aspects of PSHA. In addition to their training goals, the workshops brought together existing and new information relevant to PSHA studies in the region, providing a platform for local scientists/engineers and experts from other parts of the world to interact, cooperate and discuss various pieces of the PSHA input. This report outlines the aspects of the earthquake catalogue development and PSHA project that were specifically undertaken to form the basis for the building code update in Georgia. Georgia straddles the Greater Caucasus and Lesser Caucasus Fold and Thrust belts, where the convergent tectonic setting generates significant earthquake activity. To the east of the country, evidence indicates the possibility of a remnant subducting slab that causes earthquakes deeper than 40km. This study systematically compiles and improves all available data on local seismicity, active faults, and ground motion attenuation characteristics of the region; and builds a framework to enable a contemporary PSHA to be carried out with the engagement of local scientists. To this end, six regional workshops were organized in total over a three-year period of time, with participation from the United States, Canada and Hungary as well as countries in the Caucasus: Georgia, Armenia, Azerbaijan and Turkey. The workshops served multiple purposes such as training, data collection, interactions between local earth scientists and engineers, and brainstorming and knowledge exchange among local and international experts.

Geosciences↗

SOLPS-ITER validation with TCV L-mode discharges

This work presents a quantitative test of SOLPS-ITER simulations against tokamak a configuration variable (TCV) L-mode experiments. These simulations account for drifts, currents, kinetic neutrals, and carbon impurities providing the most complete edge transport simulations for TCV to date. The comparison is performed on nominally identical discharges carried out to assess the effectiveness of TCV’s divertor baffles in the framework of the European Plasma Exhaust program and employs numerous edge diagnostics providing a detailed code-experiment benchmark for TCV. The simulations show a qualitative consistency, but the quantitative differences remain, which are assessed herein. It is found that, for a given separatrix density, the simulations most notably yield a colder, and denser, divertor state with a higher divertor neutral pressure than measured.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

A New Analysis Tool Assessment for Rotordynamic Modeling of Gas Foil Bearings

Gas foil bearings offer several advantages over traditional bearing types that make them attractive for use in high-speed turbomachinery. They can operate at very high temperatures, require no lubrication supply (oil pumps, seals, etc.), exhibit very long life with no maintenance, and once operating airborne, have very low power loss. The use of gas foil bearings in high-speed turbomachinery has been accelerating in recent years, although the pace has been slow. One of the contributing factors to the slow growth has been a lack of analysis tools, benchmarked to measurements, to predict gas foil bearing behavior in rotating machinery. To address this shortcoming, NASA Glenn Research Center (GRC) has supported the development of analytical tools to predict gas foil bearing performance. One of the codes has the capability to predict rotordynamic coefficients, power loss, film thickness, structural deformation, and more. The current paper presents an assessment of the predictive capability of the code, named XLGFBTH (Texas A&M University). A test rig at GRC is used as a simulated case study to compare rotordynamic analysis using output from the code to actual rotor response as measured in the test rig. The test rig rotor is supported on two gas foil journal bearings manufactured at GRC, with all pertinent geometry disclosed. The resulting comparison shows that the rotordynamic coefficients calculated using XLGFBTH represent the dynamics of the system reasonably well, especially as they pertain to predicting critical speeds.

Howard, Samuel A.↗