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Simple/Quick metered data processing method that can be used for Army Metered Data Management System (MDMS) and Logistics Innovation Agency data, but may also be useful for other large data sets. Intended for large data sets when analyst has little information about the buildings.
This is a final report from a project funded by the U.S. Department of Energy to formulate and test a methodology for valuation of systems where transaction-based mechanisms coordinate the exchange of value between the system’s actors. Today, the principal commodity being exchanged is electrical energy, and such mechanisms are called transactive energy systems. The authors strove to lay a foundation for meaningful valuations of transactive systems in general, and transactive energy systems as a special case. The word valuation is used in many different ways. This report proposes a valuation methodology that is inclusive of many types of valuations. Many will be familiar with cost-benefit valuations, in which both costs and benefits are assessed to determine whether the assets are worth their cost. Another set of valuation methods attempt to optimize an outcome using available resources, as is the case with integrated resource planning. In the end, this report’s methodology was most influenced by and most resembles the integrated-resource-planning approach. Regardless, we wish to enforce the premise that all valuations are comparative and should clearly specify a baseline scenario. A long, annotated list of prior valuation studies and valuation methodologies that influenced this report has been appended to this report. Much research is being conducted today concerning transactive systems, but only a handful of transactive system mechanisms have been formulated and field tested. They are found to be quite diverse, and the documentation of the various mechanisms is uneven in breadth and quality. It is therefore not adequate to simply assert that a valuation scenario includes a transactive system; certain characteristics and qualities of the chosen transactive system mechanism must be defined and stated. The report lists and discusses most of the known transactive system mechanisms. It offers a set of questions that may be used to help specify important characteristics of the transactive system mechanisms, which should be conveyed along with other valuation results. A valuation methodology is proposed. Some abstraction is necessarily retained so that the methodology may be applied for the many purposes of today’s valuations and across grid, building, societal, and other domains. The report’s methodology advocates separation of operational timescales from long-term growth timescales. Operational models are defined as the models that inform impacts within the relatively short, often yearlong, operational time periods. Growth models define how the scenarios evolve from one operational period to the next (e.g., from year to year). We believe the recommended methodology is a critical step toward collaborative community platforms, where analysts and decision makers alike could contribute and borrow content within their expertise. The report then asks, what is unique about valuations when systems become coordinated by transactive systems? In answer, accurate valuations of transactive systems require careful adherence to the dynamic interaction between a system’s responsive elements and the system’s operational objectives. In every transactive system mechanism, elements respond to incentives that become revealed to them, and certain operational objectives become explicitly incentivized by the transactive system mechanism. The transactive system mechanisms define the important coupling between the responsive elements and the system’s objectives.
This was prepared for the business process improvement presentation to the Department of Energy. Los Alamos National Laboratory provides a single point of contact, the AskIT Service Desk, to address issues that impact customer productivity. At the most basic level, what customers want is for their calls to be received, to get a response from a knowledgeable analyst, and to have their issues resolved and their requests fulfilled. Providing a centralized, single point of contact service desk makes initiating technical or business support simple for the customer and improves the odds of immediately resolving the issue or correctly escalating the request to the next support level when necessary. Fulfilling customer requests through automated workflow also improves customer productivity and reduces costs. Finally, customers should be provided the option to solve their own problems through easy access to self-help resources such as frequently asked questions (FAQs) and how-to guides. To accomplish this, everyone who provides and supports services must understand how these processes and functions work together. Service providers and those who support services must “speak the same language” and share common objectives. The Associate Directorate for Business Innovation (ADBI) began the journey to improve services by selecting a known service delivery framework (Information Technology Infrastructure Library, or ITIL). From this framework, components that contribute significant business value were selected.
Quantifying uncertainty in model parameters is a challenging task for analysts. Soize has derived a method that is able to characterize both model and parameter uncertainty independently. This method is explained with the assumption that some experimental data is available, and is divided into seven steps. Monte Carlo analyses are performed to select the optimal dispersion variable to match the experimental data. Along with the nominal approach, an alternative distribution can be used along with corrections that can be utilized to expand the scope of this method. This method is one of a very few methods that can quantify uncertainty in the model form independently of the input parameters. Two examples are provided to illustrate the methodology, and example code is provided in the Appendix.
This report describes experiences gained in performing radiation transport computations with the SCEPTRE radiation transport code for System Generated ElectroMagnetic Pulse (SGEMP) applications. SCEPTRE is a complex code requiring a fairly sophisticated user to run the code effectively, so this report provides guidance for analysts interested in performing these types of calculations. One challenge in modeling coupled photon/electron transport for SGEMP is to provide a spatial mesh that is sufficiently resolved to accurately model surface charge emission and charge deposition near material interfaces. The method that has been most commonly used to date to compute cable SGEMP typically requires a sub-micron mesh size near material interfaces, which may be difficult for meshing software to provide for complex geometries. We present here an alternative method for computing cable SGEMP that appears to substantially relax this requirement. The report also investigates the effect of refining the energy mesh and increasing the order of the angular approximation to provide some guidance on determining reasonable parameters for the energy/angular approximation needed for x-ray environments. Conclusions for γ-ray environments may be quite different and will be treated in a subsequent report. In the course of the energy-mesh refinement studies, a bug in the cross-section generation software was discovered that may cause underprediction of the result by as much as an order of magnitude for the test problem studied here, when the electron energy group widths are much smaller than those for the photons. Results will be presented and compared using cross sections generated before and after the fix. We also describe adjoint modeling, which provides sensitivity of the total charge drive to the source energy and angle of incidence, which is quite useful for comparing the effect of changing the source environment and for determining most stressing angle of incidence and source energy. This report focusses on cable SGEMP applications, but many of the conclusions will be directly applicable for box Internal ElectroMagnetic Pulse (IEMP) modeling as well.
The purpose of this document is to provide a basic overview and understanding of risk assessment methodologies and tools from the literature and to assess the suitability of these methodologies and tools for cyber risk assessment. Sandia National Laboratories (SNL) performed this review in support of risk modeling activities performed for the Stakeholder Engagement and Cyber Infrastructure Resilience (SECIR) division of the Department of Homeland Security (DHS) Office of Cybersecurity and Communications (CS&C). The set of methodologies and tools covered in this document is not intended to be exhaustive; instead, it focuses on those that are commonly used in the risk assessment community. The classification of methodologies and tools was performed by a group of analysts with experience in risk analysis and cybersecurity, and the resulting analysis of alternatives has been tailored to address the needs of a cyber risk assessment.
In general, Dakota is the Consortium for Advanced Simulation of Light Water Reactors (CASL) delivery vehicle for verification, validation, and uncertainty quantification (VUQ) algorithms. It permits ready application of the VUQ methods described above to simulation codes by CASL researchers, code developers, and application engineers. More specifically, the CASL VUQ Strategy [33] prescribes the use of Predictive Capability Maturity Model (PCMM) assessments [37]. PCMM is an expert elicitation tool designed to characterize and communicate completeness of the approaches used for computational model definition, verification, validation, and uncertainty quantification associated with an intended application. Exercising a computational model with the methods in Dakota will yield, in part, evidence for a predictive capability maturity model (PCMM) assessment. Table 1.1 summarizes some key predictive maturity related activities (see details in [33]), with examples of how Dakota fits in. This manual offers CASL partners a guide to conducting Dakota-based VUQ studies for CASL problems. It motivates various classes of Dakota methods and includes examples of their use on representative application problems. On reading, a CASL analyst should understand why and how to apply Dakota to a simulation problem.
These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.
This proof-of-concept project focused on developing, testing, and validating a range of bibliometric, text analytic, and machine-learning based methods to explore the evolution of three photovoltaic (PV) technologies: Cadmium Telluride (CdTe), Dye-Sensitized solar cells (DSSC), and Multi-junction solar cells. The analytical approach to the work was inspired by previous work by the same team to measure and predict the scientific prominence of terms and entities within specific research domains. The goal was to create tools that could assist domain-knowledgeable analysts in investigating the history and path of technological developments in general, with a focus on analyzing step-function changes in performance, or “breakthroughs,” in particular. The text-analytics platform developed during this project was dubbed Helios. The project relied on computational methods for analyzing large corpora of technical documents. For this project we ingested technical documents from the following sources into Helios: Thomson Scientific Web of Science (papers), the U.S. Patent & Trademark Office (patents), the U.S. Department of Energy (technical documents), the U.S. National Science Foundation (project funding summaries), and a hand curated set of full-text documents from Thomson Scientific and other sources.
Sandia's Dakota software (available at http://dakota.sandia.gov) supports science and engineering transformation through advanced exploration of simulations. Specifically it manages and analyzes ensembles of simulations to provide broader and deeper perspective for analysts and decision makers. This enables them to enhance understanding of risk, improve products, and assess simulation credibility.
Central to protecting our nation's critical infrastructure is the development of methodologies for prioritizing action and supporting resource allocation decisions associated with risk-reduction initiatives. Toward this need a web-based risk assessment framework that promotes the anonymous sharing of results among water utilities is demonstrated. Anonymous sharing of results offers a number of potential advantages such as assistance in recognizing and correcting bias, identification of 'unknown, unknowns', self-assessment and benchmarking for the local utility, treatment of shared assets and/or threats across multiple utilities, and prioritization of actions beyond the scale of a single utility. The constructed framework was demonstrated for three water utilities. Demonstration results were then compared to risk assessment results developed using a different risk assessment application by a different set of analysts.
This report was developed by a team of national laboratory analysts over the period October 2015 to May 2016 and is part of a series of studies that provide background material to inform development of the second installment of the Quadrennial Energy Review (QER 1.2). The report focuses specifically on U.S. power sector generation. The report limits itself to the generation sector and does not address in detail parallel issues in electricity end use, transmission and distribution, markets and policy design, and other important segments. The report lists 15 key findings about energy system needs of the future.
This report provides an overview of the scenarios used in strategic futures workshops conducted at Sandia on September 21 and 29, 2016. The workshops, designed and facilitated by analysts in Center 100, used scenarios to enable thought leaders to think collectively about the changing aspects of global nuclear security and the potential implications for the US Government and Sandia National Laboratories.
The Department of Defense has a methodology for targeting nuclear weapons, and a jargon that is used to communicate between the analysts, planners, aircrews, and missile crews. The typical engineer or scientist in the Department of Energy may not have been exposed to the nuclear weapons targeting terms and methods. This report provides an introduction to the terms and methodologies used for nuclear targeting. Its purpose is to prepare engineers and scientists to participate in wargames, exercises, and discussions with the Department of Defense. Terms such as Circular Error Probable, probability of hit and damage, damage expectancy, and the physical vulnerability system are discussed. Methods for compounding damage from multiple weapons applied to one target are presented.
This report was developed by a team of analysts at Lawrence Berkeley National Laboratory, with Argonne National Laboratory contributing the transportation section, and is a DOE EPSA product and part of a series of “baseline” reports intended to inform the second installment of the Quadrennial Energy Review (QER 1.2). QER 1.2 provides a comprehensive review of the nation’s electricity system and cover the current state and key trends related to the electricity system, including generation, transmission, distribution, grid operations and planning, and end use. The baseline reports provide an overview of elements of the electricity system. This report focuses on end uses, electricity consumption, electric energy efficiency, distributed energy resources (DERs) (such as demand response, distributed generation, and distributed storage), and evaluation, measurement, and verification (EM&V) methods for energy efficiency and DERs.
Nuclear reactors that rely upon the fission reaction have two modes of thermal energy deposition in the reactor system: neutron absorption and gamma absorption. The gamma rays are typically generated by neutron absorption reactions or during the fission process which means the primary driver of energy production is of course the neutron interaction. In conventional reactor physics methods, the gamma heating component is ignored such that the gamma absorption is forced to occur at the gamma emission site. For experimental reactor systems like EBR-II and FFTF, the placement of structural pins and assemblies internal to the core leads to problems with power heating predictions because there is no fission power source internal to the assembly to dictate a spatial distribution of the power. As part of the EBR-II support work in the 1980s, the GAMSOR code was developed to assist analysts in calculating the gamma heating. The GAMSOR code is a modified version of DIF3D and actually functions within a sequence of DIF3D calculations. The gamma flux in a conventional fission reactor system does not perturb the neutron flux and thus the gamma flux calculation can be cast as a fixed source problem given a solution to the steady state neutron flux equation. This leads to a sequence of DIF3D calculations, called the GAMSOR sequence, which involves solving the neutron flux, then the gamma flux, then combining the results to do a summary edit. In this manuscript, we go over the GAMSOR code and detail how it is put together and functions. We also discuss how to setup the GAMSOR sequence and input for each DIF3D calculation in the GAMSOR sequence. With the GAMSOR capability, users can take any valid steady state DIF3D calculation and compute the power distribution due to neutron and gamma heating. The MC2-3 code is the preferable companion code to use for generating neutron and gamma cross section data, but the GAMSOR code can accept cross section data from other sources. To further this aspect, an additional utility code was created which demonstrates how to merge the neutron and gamma cross section data together to carry out a simultaneous solve of the two systems.
The National Nuclear Security Administration’s Office of Nuclear Smuggling Detection and Deterrence (NSDD) has facilitated the installation of more than 3,500 radiation portal monitors (RPMs) at 606 sites in 56 countries worldwide. This collection of RPMs represents the world’s largest network of radiation detectors and provides one element in the defense-in-depth approach that supports the Global Nuclear Detection Architecture. These systems support NSDD’s mission to build partner country capability to deter, detect, and interdict the illicit transport of radiological and fissile material through strategic points of entry and exit at seaports, airports, and border crossings. NSDD works collaboratively with partner countries and international organizations to optimize the operation of these RPMs. The large amount of data provided by NSDD partner countries highlights the close cooperation and partnerships NSDD has built with 56 countries around the world. Thirty-seven of these countries shared at least some RPM-related data with NSDD in fiscal year 2016. This significant level of data sharing is a key element that distinguishes the NSDD office as unique among nuclear nonproliferation programs and initiatives: NSDD can provide specific, objective, data-driven decisions and support for sustaining the radiation detection systems it helped deploy. This data analysis report summarizes and aggregates the RPM data provided to the NSDD office for analysis and review in fiscal year 2016. The data can be used to describe RPM performance and characterize the wide diversity of NSDD deployment sites. For example, NSDD deploys detector systems across sites with natural background radiation levels that can vary by a factor of approximately six from site to site. Some lanes have few occupancies, whereas others have approximately 8,000 occupancies per day and the different types of cargo that travel through a site can result in site-wide alarm rates that range from near 0% at some sites to above 3% at others. Based on the data received, the global average uptime for NSDD RPMs was above 96% for fiscal year 2016. NSDD takes all of these factors into account in making recommendations to partner countries on how to most effectively manage and maintain site operations. NSDD utilizes reports and other information products created by data analysts to efficiently allocate the resources needed to detect and ultimately interdict illicit nuclear and radiological material.
PLTEMP/ANL V4.2 is a program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of codes originally used for plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over several decades. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates or tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as onset-of-nucleate boiling ratio(ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.