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At least 1,513 records · Page 84

DIVA V1.0

The DIVA software interfaces a process in which researchers design their DNA with a web-based graphical user interface, submit their designs to a central queue, and a few weeks later receive their sequence-verified clonal constructs. Each researcher independently designs the DNA to be constructed with a web-based BioCAD tool, and presses a button to submit their designs to a central queue. Researchers have web-based access to their DNA design queues, and can track the progress of their submitted designs as they progress from "evaluation", to "waiting for reagents", to "in progress", to "complete". Researchers access their completed constructs through the central DNA repository. Along the way, all DNA construction success/failure rates are captured in a central database. Once a design has been submitted to the queue, a small number of dedicated staff evaluate the design for feasibility and provide feedback to the responsible researcher if the design is either unreasonable (e.g., encompasses a combinatorial library of a billion constructs) or small design changes could significantly facilitate the downstream implementation process. The dedicated staff then use DNA assembly design automation software to optimize the DNA construction process for the design, leveraging existing parts from the DNA repository where possible and ordering synthetic DNA where necessary. SynTrack software manages the physical locations and availability of the various requisite reagents and process inputs (e.g., DNA templates). Once all requisite process inputs are available, the design progresses from "waiting for reagents" to "in progress" in the design queue. Human-readable and machine-parseable DNA construction protocols output by the DNA assembly design automation software are then executed by the dedicated staff exploiting lab automation devices wherever possible. Since the all employed DNA construction methods are sequence-agnostic, standardized (utilize the same enzymatic master mixes and reaction conditions), completely independent DNA construction tasks can be aggregated into the same multi-well plates and pursued in parallel. The resulting sets of cloned constructs can then be screened by high-throughput next-gen sequencing platforms for sequence correctness. A combination of long read-length (e.g., PacBio) and paired-end read platforms (e.g., Illumina) would be exploited depending the particular task at hand (e.g., PacBio might be sufficient to screen a set of pooled constructs with significant gene divergence). Post sequence verification, designs for which at least one correct clone was identified will progress to a "complete" status, while designs for which no correct clones wereidentified will progress to a "failure" status. Depending on the failure mode (e.g., no transformants), and how many prior attempts/variations of assembly protocol have been already made for a given design, subsequent attempts may be made or the design can progress to a "permanent failure" state. All success and failure rate information will be captured during the process, including at which stage a given clonal construction procedure failed (e.g., no PCR product) and what the exact failure was (e.g. assembly piece 2 missing). This success/failure rate data can be leveraged to refine the DNA assembly design process.

Source record↗

DIVA V2.0

The DIVA software interfaces a process in which researchers design their DNA with a web-based graphical user interface, submit their designs to a central queue, and a few weeks later receive their sequence-verified clonal constructs. Each researcher independently designs the DNA to be constructed with a web-based BioCAD tool, and presses a button to submit their designs to a central queue. Researchers have web-based access to their DNA design queues, and can track the progress of their submitted designs as they progress from "evaluation", to "waiting for reagents", to "in progress", to "complete". Researchers access their completed constructs through the central DNA repository. Along the way, all DNA construction success/failure rates are captured in a central database. Once a design has been submitted to the queue, a small number of dedicated staff evaluate the design for feasibility and provide feedback to the responsible researcher if the design is either unreasonable (e.g., encompasses a combinatorial library of a billion constructs) or small design changes could significantly facilitate the downstream implementation process. The dedicated staff then use DNA assembly design automation software to optimize the DNA construction process for the design, leveraging existing parts from the DNA repository where possible and ordering synthetic DNA where necessary. SynTrack software manages the physical locations and availability of the various requisite reagents and process inputs (e.g., DNA templates). Once all requisite process inputs are available, the design progresses from "waiting for reagents" to "in progress" in the design queue. Human-readable and machine-parseable DNA construction protocols output by the DNA assembly design automation software are then executed by the dedicated staff exploiting lab automation devices wherever possible. Since the all employed DNA construction methods are sequence-agnostic, standardized (utilize the same enzymatic master mixes and reaction conditions), completely independent DNA construction tasks can be aggregated into the same multi-well plates and pursued in parallel. The resulting sets of cloned constructs can then be screened by high-throughput next-gen sequencing platforms for sequence correctness. A combination of long read-length (e.g., PacBio) and paired-end read platforms (e.g., Illumina) would be exploited depending the particular task at hand (e.g., PacBio might be sufficient to screen a set of pooled constructs with significant gene divergence). Post sequence verification, designs for which at least one correct clone was identified will progress to a "complete" status, while designs for which no correct clones wereidentified will progress to a "failure" status. Depending on the failure mode (e.g., no transformants), and how many prior attempts/variations of assembly protocol have been already made for a given design, subsequent attempts may be made or the design can progress to a "permanent failure" state. All success and failure rate information will be captured during the process, including at which stage a given clonal construction procedure failed (e.g., no PCR product) and what the exact failure was (e.g. assembly piece 2 missing). This success/failure rate data can be leveraged to refine the DNA assembly design process.

CHEN, JOANNA↗

2016 Annual Technology Baseline (ATB)

Consistent cost and performance data for various electricity generation technologies can be difficult to find and may change frequently for certain technologies. With the Annual Technology Baseline (ATB), National Renewable Energy Laboratory provides an organized and centralized dataset that was reviewed by internal and external experts. It uses the best information from the Department of Energy laboratory's renewable energy analysts and Energy Information Administration information for conventional technologies. The ATB will be updated annually in order to provide an up-to-date repository of current and future cost and performance data. Going forward, we plan to revise and refine the values using best available information. The ATB includes both a presentation with notes (PDF) and an associated Excel Workbook. The ATB includes the following electricity generation technologies: land-based wind; offshore wind; utility-scale solar PV; concentrating solar power; geothermal power; hydropower plants (upgrades to existing facilities, powering non-powered dams, and new stream-reach development); conventional coal; coal with carbon capture and sequestration; integrated gasification combined cycle coal; natural gas combustion turbines; natural gas combined cycle; conventional biopower. Nuclear laboratory's renewable energy analysts and Energy Information Administration information for conventional technologies. The ATB will be updated annually in order to provide an up-to-date repository of current and future cost and performance data. Going forward, we plan to revise and refine the values using best available information.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗

DoSSiER: Database of scientific simulation and experimental results

The Geant4, GeantV and GENIE collaborations regularly perform validation and regression tests for simulation results. DoSSiER (Database of Scientific Simulation and Experimental Results) is being developed as a central repository to store the simulation results as well as the experimental data used for validation. DoSSiER can be easily accessed via a web application. In addition, a web service allows for programmatic access to the repository to extract records in json or xml exchange formats. In this paper, we describe the functionality and the current status of various components of DoSSiER as well as the technology choices we made.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗

Provenance Storage, Querying, and Visualization in PBase

We present PBase, a repository for scientific workflows and their corresponding provenance information that facilitates the sharing of experiments among the scientific community. PBase is interoperable since it uses ProvONE, a standard provenance model for scientific workflows. Workflows and traces are stored in RDF, and with the support of SPARQL and the tree cover encoding, the repository provides a scalable infrastructure for querying the provenance data. Furthermore, through its user interface, it is possible to: visualize workflows and execution traces; visualize reachability relations within these traces; issue SPARQL queries; and visualize query results.

Kianmajd, Parisa↗

Storage, transportation and disposal system for used nuclear fuel assemblies

An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Storage, transportation and disposal system for used nuclear fuel assemblies

An integrated storage, transportation and disposal system for used fuel assemblies is provided. The system includes a plurality of sealed canisters and a cask sized to receive the sealed canisters in side by side relationship. The plurality of sealed canisters include an internal basket structure to receive a plurality of used fuel assemblies. The internal basket structure includes a plurality of radiation-absorbing panels and a plurality of hemispherical ribs generally perpendicular to the canister sidewall. The sealed canisters are received within the cask for storage and transportation and are removed from the cask for disposal at a designated repository. The system of the present invention allows the handling of sealed canisters separately or collectively, while allowing storage and transportation of high burnup fuel and damaged fuel to the designated repository.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Chemical Pre-Treatment of used Fuel for Long–Term Storage

Repository programs throughout the world have been slowed by the need for increased local public involvement in the siting and licensing process. The result has been an increase in the dry storage of used fuel at reactor sites and the potential that such storage may be extended for many decades, even centuries. While there are sound technical reasons to believe that dry storage can be conducted safely, there are increasing concerns that the ultimate transfer to either a future repository or a centralized separations plant may result in fractured cladding and serious handling issues, including criticality concerns. These concerns would be increased for higher burn-up fuels. Currently, various chemical pre-treatment processes under R&D for application to commercial used oxide fuel have been investigated at the laboratory scale as methods to simplify and increase the safety of the remaining stages of conventional solvent extraction processing. This includes advanced decladding methods and various oxidation/reduction processes designed to release volatile and semi-volatile fission products, produce finely divided uranium oxide powder, and ameliorate the subsequent nitric acid dissolution step. The paper examines the potential for combining several chemical and physical pre-treatment steps to minimize long-term concerns about safe transport of used fuel, possibly providing another option for future nuclear waste management. Laboratory data from both cold and hot testing will provide the basis for the evaluation. An example of a potential pre-treatment process includes shearing, advanced voloxidation and off-gas treatment, the possible mixing of the resulting uranium oxide with a secondary oxide, and densification and recanning in nitric acid-soluble storage containers for extended time periods. Chemical decladding may be feasible to replace shearing. Zirconium recycle may also be feasible, significantly reducing high level waste quantities. Both analytic and experimental data will be applied to the examination of this potential fuel cycle option.

Bresee, James↗

Simulation of Uranium(VI) Speciation and Sorption in Montmorillonite Clay in the Presence of Calcite Impurities

Montmorillonite is an abundant clay mineral that has been identified as the primary component in engineered bentonite barrier systems in future nuclear waste repositories. Selective sorption onto planar and edge surfaces of clay particles make it a great candidate for the attentuation of uranium(VI). Although previous works have investigated uranium(VI) interactions with this clay, few have done so in the presence of calcite, a mineral impurity in bentonite that is understood to alter U(VI) speciation, sorption, and consequently its mobility. PHREEQC software is used to simulate the chemical solution speciation of U(VI) under various known repository conditions, in the absence and presence of calcite, and across a range of U(VI)concentrations (1.0E-7 to 1.0E 5 M), and partial pressures of CO 2 (zero to 2% CO2) (Fig. 1). In addition, the kinetic dissolution of calcite is simulated to gain a better understanding of how quickly chemical equilibrium can be reached in these systems. Lastly, surface complexation models will be developed to characterize the influence of calcite impurities on the formation of ternary U(VI)-Ca-carbonate solution species on U(VI) sorption behavior. Our results indiciate that the presence of Ca 2+ and CO 3 2- ions released from calcite leads to U(VI)-Ca-carbonate species with a negative or neutral charge even without the presence of CO 2 . Quantification of these species will allow for the development of more accurate U(VI)-montmorillonite sorption models. Calcite precipitation at alkaline pH further complicates speciation by creating a competation for Ca 2+ . This research is being performed using funding received from the DOE Office of Nuclear Energy’s NEUP.

Pistorino, Jonathan C.↗

Codisposal Waste Package Loading Options for DOE SNF and HLW PPT

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. DOE must continue to safely store that SNF, transport it to an interim storage site or a repository, and dispose of it. These fuels come from a wide range of reactor types with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties, resulting in a large diversity of reactor and fuel designs. Because of the wide variety and conditions of SNF, a robust DOE Standard Canister was proposed that would confine radionuclides and preclude moderator. The DOE Standard Canister had four variations: 3.05-meter (10-foot) or 4.57-meter (15-foot) length, and 45.7-cm (18-inch) or 61.0-cm (24-inch) diameter. For ultimate disposal in the Yucca Mountain Repository, these canisters were to be grouped with 61.0-cm (24-inch) diameter high level waste (HLW) canisters in a 2.13-meter (84-inch) diameter co-disposal waste package. The smaller 45.7-cm (18-inch) diameter DOE Standard Canister could be placed in the middle of five HLW canisters. The larger 61.0-cm (24-inch) diameter DOE Standard Canister would take the place of one of the five HLW canisters on the outer ring in the co-disposal waste package. No DOE Standard Canisters have been loaded. A preliminary evaluation has estimated the number of elements of a fuel type that can fit into the different sizes of the DOE Standard Canister, but no definitive loading configuration has been selected. Changing the loading configuration could impact the number of loadable DOE Standard Canisters and the number of co-disposal waste packages needed for eventual disposition. This paper conveys the ranges of DOE Standard Canisters and HLW canisters that may be produced under certain conditions. It also examines the differences in the estimated number of co-disposal waste packages produced for eventual disposal when using different loading strategies in the DOE Standard Canister for Advanced Test Reactor (ATR), Peach Bottom, and High Flux Isotope Reactor (HFIR) SNF. Changing the loading configurations of ATR, Peach Bottom, and HFIR SNF slightly impacted the number of co-disposal waste packages that may be needed for ultimate disposal. The change in loading configuration was more impactful when a different canister was used, as opposed to varying the number of elements that could fit inside the same size canister. In one case, a reduction of co-disposal waste packages could be achieved by allowing mixing of short HLW canisters with long DOE Standard Canisters. The main conclusion from this analysis is that the ratio between HLW canisters and DOE Standard Canisters will drive the total number of co-disposal waste packages. If too many HLW canisters (i.e., more than five times the number of 18-inch DOE standard canisters) or DOE Standard Canisters are produced, some co-disposal waste packages may not have all positions filled. A co-disposal waste package may be filled with all HLW with no DOE Standard Canister, or a co-disposal waste package could be filled with a single DOE Standard Canister. A ratio that does not closely align to optimum could allow for the design of waste packages that hold just HLW canisters or just DOE SNF canisters.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Codisposal Waste Package Loading Options for DOE SNF and HLW

The U.S. Department of Energy (DOE) is responsible for managing spent nuclear fuel (SNF) that is currently in, or will in the future come into, its possession. DOE must continue to safely store that SNF, transport it to an interim storage site or a repository, and dispose of it. These fuels come from a wide range of reactor types, such as light- and heavy-water-moderated reactors, graphite-moderated reactors, and breeder reactors, with various cladding materials and enrichments. Many of these reactors, now decommissioned, had unique design features, such as core configuration, fuel element and assembly geometry, moderator and coolant materials, operational characteristics, and neutron spatial and spectral properties resulting in a large diversity of reactor and fuel designs. Because of the wide variety and condition of SNF, a robust canister termed the DOE Standardized SNF Canister was proposed that would confine radionuclides and preclude moderator. The DOE Standard SNF Canister had four variations consisting of a 10- or 15-foot length and an 18- or 24-inch diameter. For ultimate disposal in the Yucca Mountain Repository these canisters were to be grouped with 24-inch diameter HLW canisters in a codisposal waste package. The 18-inch DOE Standardized SNF Canister could be placed in the middle of five HLW canisters. The 24-inch DOE Standardized SNF Canister would take the place of one the five HLW canister on the outer ring in the codisposal waste package. No DOE Standardized SNF Canisters have been loaded, and many fuel types can be loaded in multiple canisters. Depending on the selected diameter of the DOE Standardized SNF Canister for a particular fuel, the number of potential codisposal waste packages could vary. In addition, different types of DOE SNF may be able to be combined in a single canister. This could significantly change the number of DOE SNF canisters that are loaded and result in a decrease of the number of total potential codisposal waste packages. Ultimately, if too many HLW canisters need disposition, then a reduction in the number of DOE Standardized SNF Canisters will change the final total of codisposal waste packages very little. Likewise, if the ratio of DOE Standardized SNF Canisters to HLW canister produced is too high, then DOE Standardized SNF Canisters will not have enough HLW canisters to surround them in the codisposal waste package. This paper examines the results of differing loading strategies for a few DOE SNFs. It also compares the ranges of DOE Standardized SNF Canisters that may be produced to the ranges of HLW canisters that may be produced in order to determine potential scenarios.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Standardized Architecture for a Mega-Biobank Phenomic Library: The Million Veteran Program (MVP)

Electronic health records (EHRs) provide a wealth of data for phenotype development in population health studies, and researchers invest considerable time to curate data elements and validate disease definitions. The ability to reproduce well-defined phenotypes increases data quality, comparability of results and expedites research. In this paper, we present a standardized approach to organize and capture phenotype definitions, resulting in the creation of an open, online repository of phenotypes. This resource captures phenotype development, provenance and process from the Million Veteran Program, a national mega-biobank embedded in the Veterans Health Administration (VHA). To ensure that the repository is searchable, extendable, and sustainable, it is necessary to develop both a proper digital catalog architecture and underlying metadata infrastructure to enable effective management of the data fields required to define each phenotype. Our methods provide a resource for VHA investigators and a roadmap for researchers interested in standardizing their phenotype definitions to increase portability.

Knight, Kathryn↗

Reply to: How much does corrosion of nuclear waste matrices matter

We thank Mallants and Chapman for their comments on our recent work and are mostly in agreement. However, while the applicability of the results to different environmental conditions requires more attention, we maintain that this study could be relevant even to repositories in saturated reduced environments. Oxidants other than oxygen might be present in anoxic groundwater repository environments and could also trigger the corrosion mechanism we described. .

Guo, Xiaolei↗

Hybrid User Manual

This document is the user manual for the Hybrid Repository that is a part of the Integrated Energy Systems program. The Hybrid Repository is a collection of physical modelica models used for the creation of transient integrated energy systems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Fuel Matrix Degradation Model Development Update: Alloy Corrosion Rates and Hydrogen Generation

This work is being performed as part of the DOE NE Spent Fuel and Waste Science and Technology Campaign, Argillite and Crystalline Rock R&D work packages: SF-19AN01030101 and SF-19AN01030201. This document meets the milestone M4SF-19AN010301013 for Argillite R&D and the milestone M4SF-19AN010302013 for Crystalline R&D. The fuel matrix degradation (FMD) model calculates the degradation rate of spent UO2 fuel based on fundamental electrochemical theory and thermodynamics. The model provides the quantitative basis for radionuclide source term predictions in the generic disposal system analysis (GDSA) performance assessment (PA) model. The FMD model has been implemented in a manner that facilitates its integration with the GDSA-PA code and a preliminary, integrated FMD-GDSA model has been successfully tested (Jerden et al., 2017). The specific focus of on-going work is to accurately and quantitatively represent the generation of H2 in a breached waste package and its effect on the spent fuel degradation rate so that this key process can be accurately represented in PA models. It has been shown experimentally that millimolar concentrations of dissolved H2 in contact with spent fuel will inhibit oxidative dissolution and decrease the fuel degradation rates by as much as three orders of magnitude (e.g., Röllin et al., 2001, Ollila, 2008). This will lead to a significant decrease in the radionuclide source term values used in repository PA models. Sensitivity studies using the FMD model have shown that the dissolved H2 concentration is the dominant environmental variable affecting the UO2 spent fuel dissolution rate as long as a source of H2 persists in the breached waste package (Jerden et al., 2015). Therefore, on-going experimental and modeling efforts focus on H2 production, accumulation and interfacial reactions within a breached waste package. The anoxic corrosion of metallic engineering materials within a breached waste package will be the main source of H2 in crystalline and argillite rock repository systems. The most important H2-producing alloys present within a waste package include stainless steel, carbon steel, and aluminum alloys and the Zircaloy fuel cladding. A steel corrosion module was added to the FMD model to account for these H2 sources and couple the H2 generation rate with fuel degradation processes (Jerden et al., 2017). As discussed below, there is a need for experimental data from electrochemical corrosion tests to quantify the H2 generation rates due to the corrosion of these waste package alloys as a function of key environmental variables (Eh, pH, groundwater chemistry). This report presents results from series of electrochemical corrosion tests with materials present in waste packages that provide information needed to parameterize and validate the FMD model. The alloys tested include 316L stainless steel (316SS), AISI 4320 carbon steel (C-steel) and Zircaloy-4 fuel cladding (unirradiated). Potentiodynamic polarization and potentiostatic polarization tests were performed in buffer solutions at pH 4, 7 and 10 with NaCl concentrations of 0, 4 and 100 millimolar (mM). The potentiostatic tests were performed at voltages varying from 0.5 volts down to -0.66 volts, vs the standard hydrogen electrode (SHE), and all tests were performed at laboratory ambient temperatures (~22 °C). The corrosion rates measured during the electrochemical tests were used as direct input to a new FMD/in-package chemistry model that is being developed to provide accurate long-term spent fuel degradation rates as chemical conditions (Eh, pH, speciation) evolve within a breached waste package. This new model was developed by coupling the FMD model with the reactive transport code X1t, which is a module within the Geochemist’s Workbench (GWB) reaction path modeling code (Jerden et al., 2019). The reactive transport model consists of a 1D domain discretized with 21 reaction diffusion cells and includes a single waste packag

Jerden, James↗

Hybrid User Manual

This document is the user manual for the Hybrid Repository that is a part of the Integrated Energy Systems program. The Hybrid Repository is a collection of physical modelica models used for the creation of transient integrated energy systems.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Engineered barrier material interactions at elevated temperatures: Bentonite-metal interactions under elevated temperature conditions

The development of deep repository concepts in the USA is evaluating generic options for disposal of heat-generating nuclear spent nuclear fuel (SNF) waste for a suite of host rock media (clay rock, granite, salt). Large waste canister designs (e.g., dual purpose canisters or DPC’s) are currently being considered to accommodate many SNF assemblies. Large amounts of SNF in the waste packages will produce high thermal loads generating temperatures in excess of 200 °C for long periods of time. For a bentonite backfilled repository concept, prolonged exposure to high temperatures will induce chemical reactions in the engineered barrier system (EBS), particularly at barrier interfaces. Our focus is on experimental investigations and the application of thermodynamic modeling to evaluate clay-zeolite phase equilibria as a function of temperature and fluid chemistry. Experimental work on barrier material interactions under hydrothermal conditions (150 – 300 °C, 15-16 MPa) has elucidated mineral phase changes in Wyoming (Colony mine) bentonite in the presence of steel phases. Glassy material in bentonite is replaced by analcime-wairakite phases, and through apparent clinoptilolite recrystallization. The initial increase in dissolved silica leads to authigenic quartz formation. Such mineral assemblage suggests an initially silica-rich environment (analcime, clinoptilolite) moving towards Si-depleted conditions. Analcime-wairakite compositions suggest a well-defined solid-solution between these Na and Ca end-members. Smectite clay in these experiments is stable with Fe-saponite and chlorite growth co-existing with binary/ternary sulfides at steel interfaces. Little or no illite was observed in the reaction products which could be tied to silica oversaturation and low K in the system. The thermodynamic analysis is used to evaluate thermodynamic data and develop phase diagrams to describe stability field relations of secondary mineral phase occurrences. This analysis allows for delineation of potential reaction pathways in bentonite clay degradation and interactions with metallic phases. One example of this is the important role of dissolved silica plus other phase components to the formation of alteration mineralogy observed in these interactions. All these investigations are key to the assessment of thermally-induced degradation zones in the EBS during the thermal period and their effect on barrier performance in the safety assessment.

Jove-Colon, Carlos↗

Standardized Architecture for a Mega-Biobank Phenomic Library: The Million Veteran Program (MVP)

Electronic health records (EHRs) provide a wealth of data for phenotype development in population health studies, and researchers invest considerable time to curate data elements and validate disease definitions. The ability to reproduce well-defined phenotypes increases data quality, comparability of results and expedites research. In this paper, we present a standardized approach to organize and capture phenotype definitions, resulting in the creation of an open, online repository of phenotypes. This resource captures phenotype development, provenance and process from the Million Veteran Program, a national mega-biobank embedded in the Veterans Health Administration (VHA). To ensure that the repository is searchable, extendable, and sustainable, it is necessary to develop both a proper digital catalog architecture and underlying metadata infrastructure to enable effective management of the data fields required to define each phenotype. Our methods provide a resource for VHA investigators and a roadmap for researchers interested in standardizing their phenotype definitions to increase portability.

Knight, Kathryn↗