SEARCH · Engineering Papers

Results for “consequence”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

SourceResearch connectionFrom yearThrough year

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,765 records · Page 98

All sources NASA NTRS DOE OSTI

DOE OSTI2013

H→γγ as a Triangle Anomaly: Possible Implications for the Hierarchy Problem

The Standard Model calculation of H→γγ has the curious feature of being finite but regulator-dependent. While dimensional regularization yields a result which respects the electromagnetic Ward identities, additional terms which violate gauge invariance arise if the calculation is done setting d = 4. This discrepancy between the d=4 – ϵ and d = 4 results is recognized as a true ambiguity which must be resolved using physics input; as dimensional regularization respects gauge invariance, the d = 4 – ϵ calculation is accepted as the correct SM result. However, here we point out another possibility; working in analogy with the gauge chiral anomaly, we note that it is possible that the individual diagrams do violate the electromagnetic Ward identities, but that the gauge-invariance-violating terms cancel when all contributions to H→γγ, both from the SM and from new physics, are included. We thus examine the consequences of the hypothesis that the d = 4 calculation is valid, but that such a cancellation occurs. We work in general renormalizable gauge, thus avoiding issues with momentum routing ambiguities. We point out that the gauge-invariance-violating terms in d = 4 arise not just for the diagram containing a SM $W^{\pm}$ boson, but also for general fermion and scalar loops, and relate these terms to a lack of shift invariance in Higgs tadpole diagrams. We then derive the analogue of "anomaly cancellation conditions", and find consequences for solutions to the hierarchy problem. In particular, we find that supersymmetry obeys these conditions, even if it is softly broken at an arbitrarily high scale.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

DOE OSTI2012

The SAS4A/SASSYS-1 Safety Analysis Code System

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic safety analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This software can be used to assess the safety of a prescribed reactor design, but it cannot be used to configure a design to meet targeted performance objectives. Detailed, mechanistic models of steady-state and transient thermal,hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems and the balance-of-plant to accidents caused by changes in coolantflow, oss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models In SAS4A/SASSYS-1 were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options including lead and LBE. METHOD OF SOLUTION: Each subassembly in the reactor core is represented with single or multiple-pin models. One channel represents one or more fuel pins and many channels are employed for a whole-core representation. Heat transfer in each pin is modeled with a two-dimensional (r/z) heat conduction equation. Single and two-phase coolant thermal-hydraulics are simulated with a unique, one-dimensional (axial) multiple-bubble liquid metal boiling model. The transient fuel and cladding mechanical oehavior models are integrated with fission product production,release, and transport models. UNIQUE FEATURES: The physical models in SAS4AJSASSYS-1 are highly detailed 1umerical representations of reactor accident conditions based on extensive laboratory and test reactor results. The models are specialized to liquid metal (sodium) cooled fast reactors with oxide or metallic fuel clad with stainless steel.

Source record↗

DOE OSTI2016

PRELIMINARY TESTING OF NANOPARTICLE EFFECTIVENESS IN BINDER JETTING APPLICATIONS

Binder jetting works by selectively depositing a binder with an inkjet print head into layers of powdered material. Compared with other metal Additive Manufacturing (AM) processes, binder jetting has significant potential for near-term adoption in manufacturing environments due to its reliability and throughput. The Achilles heel of binder jetting, however, is the inability to produce fully dense, single-alloy materials. The lack of density in printed binder jet parts is strictly dictated by the packing factor of the powder feedstock. Adding nanoparticles during printing will not only increase the part s packing factor but may also serve as a sintering aid. This study focuses on the effect of both the binder and nanoparticles on the final part density. As an unintended consequence of high nanoparticle loading, printed parts underwent a significant increase in porosity during the curing process. This unintended consequence is the apparent result of the nanoparticles blocking the exit of the solvent vapor during the curing step. Additionally, nanoparticle use for densification is validated with SEM imagery.

Bailey, Alta C↗

DOE OSTI2015

Calibrating Nonlinear Soil Material Properties for Seismic Analysis Using Soil Material Properties Intended for Linear Analysis

Seismic analysis of nuclear structures is routinely performed using guidance provided in “Seismic Analysis of Safety-Related Nuclear Structures and Commentary (ASCE 4, 1998).” This document, which is currently under revision, provides detailed guidance on linear seismic soil-structure-interaction (SSI) analysis of nuclear structures. To accommodate the linear analysis, soil material properties are typically developed as shear modulus and damping ratio versus cyclic shear strain amplitude. A new Appendix in ASCE 4-2014 (draft) is being added to provide guidance for nonlinear time domain SSI analysis. To accommodate the nonlinear analysis, a more appropriate form of the soil material properties includes shear stress and energy absorbed per cycle versus shear strain. Ideally, nonlinear soil model material properties would be established with soil testing appropriate for the nonlinear constitutive model being used. However, much of the soil testing done for SSI analysis is performed for use with linear analysis techniques. Consequently, a method is described in this paper that uses soil test data intended for linear analysis to develop nonlinear soil material properties. To produce nonlinear material properties that are equivalent to the linear material properties, the linear and nonlinear model hysteresis loops are considered. For equivalent material properties, the shear stress at peak shear strain and energy absorbed per cycle should match when comparing the linear and nonlinear model hysteresis loops. Consequently, nonlinear material properties are selected based on these criteria.

nonlinear↗

DOE OSTI2016

Reservoir response to thermal and high-pressure well stimulation efforts at Raft River, Idaho

An injection stimulation test begun at the Raft River geothermal reservoir in June, 2013 has produced a wealth of data describing well and reservoir response via high-resolution temperature logging and distributed temperature sensing, seismic monitoring, periodic borehole televiewer logging, periodic stepped flow rate tests and tracer injections before and after stimulation efforts. One of the primary measures of response to the stimulation is the relationship between fluid pressure and flow rate, short-term during forced flow rate changes and the long-term change in injectivity. In this paper we examine that hydraulic response using standard pumping test analysis methods, largely because pressure response to the stimulation was not detected, or measurable, in other wells. Analysis of stepped rate flow tests supports the inference from other data that a large fracture, with a radial extent of one to several meters, intersects the well in the target reservoir, suggests that the flow regime is radial to a distance of only several meters and demonstrates that the pressure build-up cone reaches an effective constant head at that distance. The well’s longer term hydraulic response demonstrated continually increasing injectivity but at a dramatically faster rate later from ~2 years out and continuing to the present. The net change in injectivity is significantly greater than observed in other longterm injectivity monitoring studies, with an approximately 150–fold increase occurring over ~2.5 years. While gradually increasing injectivity is a likely consequence of slow migration of a cooling front, and consequent dilation of fractures, the steady, ongoing, rate of increase is contrary to what would be expected in a radial or linear flow regime, where the cooling front would slow with time. As a result, occasional step-like changes in injectivity, immediately following high-flow rate tests suggest that hydro shearing during high-pressure testing altered the near-well permeability structure.

15 GEOTHERMAL ENERGY↗

DOE OSTI2015

ARAM-W Automated Risk Assessment Methodology for Water and Wastewater

The Automated Risk Assessment Methodology for Water and Wastewater Utilities, ARAM-W, is a systematic, risk-based approach used to evaluate water and wastewater utilities from a range of threats that could cause an undesired event. It leverages the automated RAM-CI (critical infrastructure) prototype tool completed in September 2008, the manual RC RAM-W (RAMCAP compliant) approach completed in September 2007, the manual RAM-W, version 2, approach completed in 2002, and the requirements and needs of the Water Sector, EPA and DHS. The ARAM-W tool is a user-friendly tool which provides a sound scientific and technical framework to look at the total protection system. The ARAM-W tool provides the user the ability to determine a relative risk based on the threat, consequences and protection system effectiveness/vulnerability. It is a stand-alone PC-based tool which is compatible with the NIPP risk assessment methodology criteria as well as RAMPCAP. The ARAM-W process provides through a series of steps the ability to plan for the risk assessment; characterize the utility, facilities and assets; define the threat; identify undesired events and consequences; evaluate protection system effectiveness; determine risk; propose and evaluate upgrades if the risk is not acceptable and report on the assessment results.

Dang, Tam↗

DOE OSTI2017

SAS4A/SASSYS-1 VERSION 5.2 SAFETY ANALYSIS CODE SYSTEM

SAS4A/SASSYS-1 is a software simulation tool used to perform deterministic safety analysis of anticipated events as well as design basis and beyond design basis accidents for advanced nuclear reactors. This software can be used to assess the safety of a prescribed reactor design, but it cannot be used to configure a design to meet targeted performance objectives. Detailed, mechanistic models of steady-state and transient thermal, hydraulic, kinetic, and mechanical phenomena are employed to describe the response of the reactor core, the reactor primary and secondary coolant loops, the reactor control and protection systems and the balance-of-plant to accidents caused by changes in coolant flow, oss of heat rejection, or reactivity insertion. The consequences of single and double-fault accidents can be modeled, including fuel and coolant heating, fuel and cladding mechanical behavior, core reactivity feedbacks, coolant loop performance including natural circulation, and decay heat removal. Analyses are typically terminated upon demonstration of reactor and plant shutdown to permanently coolable conditions or upon violation of design basis margins. The objective of the analysis is to quantify accident consequences as measured by the transient behavior of system performance parameters, such as fuel and cladding temperatures, reactivity, and cladding strain. Originally developed for analysis of sodium cooled reactors with oxide fuel clad by stainless steel, the models In SAS4A/SASSYS-1 were subsequently extended and specialized to metallic fuel clad with advanced alloys and to several other coolant options including lead and LBE. METHOD OF SOLUTION: Each subassembly in the reactor core is represented with single or multiple-pin models. One channel represents one or more fuel pins and many channels are employed for a whole-core representation. Heat transfer in each pin is modeled with a two-dimensional (r/z) heat conduction equation. Single and two-phase coolant thermal-hydraulics are simulated with a unique, one-dimensional (axial) multiple-bubble liquid metal boiling model. The transient fuel and cladding mechanical behavior models are integrated with fission product production,release, and transport models. UNIQUE FEATURES: The physical models in SAS4AJSASSYS-1 are highly detailed 1umerical representations of reactor accident conditions based on extensive laboratory and test reactor results. The models are specialized to liquid metal (sodium) cooled fast reactors with oxide or metallic fuel clad with stainless steel.«

Fanning, Thomas↗

DOE OSTI2017

A Methodology for Dynamic Security Risk Quantification and Optimal Resource Allocation of Security Assets

In this presentation we overview a methodology for dynamic security risk quantification and optimal resource allocation of security assets for high profile venues. This methodology is especially applicable to venues that require security screening operations such as mass transit (e.g., train or airport terminals), critical infrastructure protection (e.g., government buildings), and largescale public events (e.g., concerts or professional sports). The method starts by decomposing the three core components of risk -- threat, vulnerability, and consequence -- into their various subcomponents. For instance, vulnerability can be decomposed into availability, accessibility, organic security, and target hardness and each of these can be evaluated against the potential threats of interest for the given venue. Once evaluated, these subcomponents are rolled back up to compute the specific value for the vulnerability core risk component. Likewise, the same is done for consequence and threat, and then risk is computed as the product of these three components. A key aspect of our methodology is dynamically quantifying risk. That is, we incorporate the ability to uniquely allow the subcomponents and core components, and in turn, risk, to be quantified as a continuous function of time throughout the day, week, month, or year as appropriate.

Brigantic, Robert T.↗

DOE OSTI2018

Framework and methodology for supply chain lifecycle analytics

The various technologies presented herein relate to pertaining to identifying and mitigating risks and attacks on a supply chain. A computer-implemented representation of a supply chain is generated comprising nodes (locations) and edges (objects, information). Risk to attack and different attack vectors can be defined for the various nodes and edges, and further, based upon the risks and attacks, (difficulty, consequence) pairs can be determined. One or more mitigations can be generated to increase a difficulty of attack and/or reduce consequence of an attack. The one or more mitigations can be constrained, e.g., by cost, time, etc., to facilitate determination of how feasible a respective mitigation is to implement with regard to finances available, duration to implement, etc. A context-free grammar can be utilized to identify one or more attacks in the supply chain. Further, the risks can undergo a ranking to enable mitigation priority to be determined.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2018

Soiling Related Abrasion and the Development of a PV Abrasion Standard

The natural soiling and consequent cleaning of photovoltaic (PV) modules result in abrasion damage to the cover glass. The durability of the front glass has important economic consequences, including the use of antireflective and/or antisoiling coatings as well as the method and frequency of cleaning-maintenance. The abrasion of glass and coated glass has been explored in a field study, including the soiling-prone locations of: Dubai (United Arab Emirates), Kuwait city (Kuwait), Mesa (Arizona), Mumbai (India), and Sacramento (California). Dry brush cleaned specimens will be compared to those subject to artificial brush testing or falling sand abrasion. The characteristics of material integrity, optical transmittance, surface energy, surface roughness, and scratch size were examined using an optical microscope, spectrophotometer, contact goniometer (for water), interferometer, and atomic force microscope, respectively. The findings of this study will be used to provide feedback regarding the cleaning equipment, cleaning methods, and coatings used in the PV industry. The study here will also be used to aid the development of an abrasion standard for the PV industry.

14 SOLAR ENERGY↗

DOE OSTI2016

Safety Analysis Challenges With Restart of the Idaho National Laboratory (INL) Transient Reactor Test (TREAT) Facility

Safety Analysis Challenges With Restart of the Idaho National Laboratory Transient Reactor Test (TREAT) Facility Dr. James Parry David Broussard Anthony LaPorta Douglas Gerstner Battelle Energy Alliance Idaho National Laboratory P.O. Box 1625/MS6194 Idaho Falls, ID 83415 The Transient Reactor Test Facility, commonly referred to as TREAT, is an air-cooled, thermal spectrum test reactor facility designed to: • Induce intense fission heating in the nuclear fuel being tested. • Test nuclear reactor fuels under severe reactor-accident conditions. • Study fuel meltdowns, metal-water reactions, interactions between overheated fuel and coolant, and the transient behavior of fuels for high-temperature systems. • Provide nondestructive test data through neutron radiography of fuel samples. • Allow for the detailed monitoring of the experiments during the test. TREAT was specifically designed to test prototypic-sized reactor fuel pins and bundles under transient power conditions. During its 35 year operating history, it conducted thousands of transient experiments. Today, these capabilities are required for predicting safety margins for next generation fuels and advanced reactor design safety. Even though TREAT has not operated since 1994, return to operable conditions and resumption of testing will occur in FY 2018. This paper discusses the unique challenges with performing and documenting the safety analyses that demonstrate that the TREAT restart may be performed safely. This includes a discussion of 1) the TREAT capability, design, and operations, 2) identification of the format and content of the required safety basis document changes, and 3) demonstration that the consequences of potential accidents during TREAT operations meet the consequence guidelines.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2017

Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

In the aftermath of the March 2011 multi-unit accident at the Fukushima Daiichi nuclear power plant (Fukushima), the nuclear community has been reassessing certain safety assumptions about nuclear reactor plant design, operations and emergency actions, particularly with respect to extreme events that might occur and that are beyond each plant’s current design basis. Because of our significant domestic investment in nuclear reactor technology (99 operating reactors in the fleet of commercial LWRs with five under construction), the United States has been a major leader internationally in these activities. The U.S. nuclear industry is voluntarily pursuing a number of additional safety initiatives. The NRC continues to evaluate and, where deemed appropriate, establish new requirements for ensuring adequate protection of public health and safety in the occurrence of low probability events at nuclear plants; (e.g., mitigation strategies for beyond design basis events initiated by external events like seismic or flooding initiators). The DOE has also played a major role in the U.S. response to the Fukushima accident. Initially, DOE worked with the Japanese and the international community to help develop a more complete understanding of the Fukushima accident progression and its consequences, and to respond to various safety concerns emerging from uncertainties about the nature of and the effects from the accident. DOE R&D activities are focused on providing scientific and technical insights, data, analyses methods that ultimately support industry efforts to enhance safety. These activities are expected to further enhance the safety performance of currently operating U.S. nuclear power plants as well as better characterize the safety performance of future U.S. plants. In pursuing this area of R&D, DOE recognizes that the commercial nuclear industry is ultimately responsible for the safe operation of licensed nuclear facilities. As such, industry is considered the primary “end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2019

Comparison of Mesoscale Model Setups for Offshore Wind Resource Assessment: A New Jersey Case Study

As the offshore wind energy industry begins to emerge on the U.S. northeast coast, there is an urgent need for an accurate characterization of the offshore wind resource. The offshore regime features various resource characteristics not seen onshore, including frequent coastal low-level jets, non-steady surface conditions (e.g. waves), complex air-sea and land-sea interactions, and coastal mesoscale circulations (e.g. sea breeze). An accurate characterization of these phenomena in mesoscale models is critical to support accurate energy production estimates and forecasting for future offshore wind farms. However, a robust validation of mesoscale model performance in the offshore U.S. has been limited due to sparse data and difficulty obtaining offshore measurements. Consequently, it is unclear how mesoscale models perform offshore and which suite of parameterizations and input data are best suited for accurate resource characterization. To begin addressing this knowledge gap, the National Renewable Energy Laboratory, in collaboration with the Rutgers Center for Ocean Observing Leadership, conducted a mesoscale model validation campaign in the offshore environment. The Weather Research and Forecasting (WRF) model was assessed and the New Jersey offshore coastal area served as the validation domain. An ensemble of WRF simulations conducted over a one-year period sampled from a range of model setups, including different reanalysis and forecast data inputs, sea surface temperature data inputs, planetary boundary layer parameterizations, and WRF versions. The performance of each ensemble member was validated against a network of observations including offshore floating lidars, coastal met towers, offshore platform measurements, and offshore buoys. Model performance was assessed on annual, seasonal, and diurnal scales with particular emphasis on the vertical wind speed profile and temperature stratification. Preliminary results from this validation work will be presented here. The key WRF inputs and parameters driving changes in the modeled wind resource will be discussed and preliminary recommendations on WRF setups for offshore wind resource modeling will be proposed. The application of this validation exercise to other proposed offshore wind energy areas in the U.S. will also be discussed. mesoscale circulations (e.g. sea breeze). An accurate characterization of these phenomena in mesoscale models is critical to support accurate energy production estimates and forecasting for future offshore wind farms. However, a robust validation of mesoscale model performance in the offshore U.S. has been limited due to sparse data and difficulty obtaining offshore measurements. Consequently, it is unclear how mesoscale models perform offshore and which suite of parameterizations and input data are best suited for accurate resource characterization. To begin addressing this knowledge gap, the National Renewable Energy Laboratory, in collaboration with the Rutgers Center for Ocean Observing Leadership, conducted a mesoscale model validation campaign in the offshore environment. The Weather Research and Forecasting (WRF) model was assessed and the New Jersey offshore coastal area served as the validation domain. An ensemble of WRF simulations conducted over a one-year period sampled from a range of model setups, including different reanalysis and forecast data inputs, sea surface temperature data inputs, planetary boundary layer parameterizations, and WRF versions. The performance of each ensemble member was validated against a network of observations including offshore floating lidars, coastal met towers, offshore platform measurements, and offshore buoys. Model performance was assessed on annual, seasonal, and diurnal scales with particular emphasis on the vertical wind speed profile and temperature stratification. Preliminary results from this validation work will be presented here. The key WRF inputs and parameters driving changes in the modeled wind resource will be discussed and preliminary recommendations on WRF setups for offshore wind resource modeling will be proposed. The application of this validation exercise to other proposed offshore wind energy areas in the U.S. will also be discussed.

17 WIND ENERGY↗

DOE OSTI2019

Effects of Plate Curvature on Thermo-Mechanical Performance of U-10Mo Monolithic Fuel Plates

Monolithic fuel is a candidate fuel form being considered for the conversion of high-performance research reactors. This plate-type fuel consists of a high-density, U-Mo fuel in a monolithic form that is sandwiched between zirconium diffusion barriers, and encapsulated in an aluminum cladding. To date, large number of plates have been irradiated with satisfactory performance. The program is now moving into the qualification phase, a predecessor to the timely conversion of the target reactors. Since each reactor employs distinct fuel plate geometries for various consideration, resulting nearly 50 distinct plate geometries with unique plate design features, a single “generic” plate geometry capturing all of the extremities is not achievable. This limitation consequently requires much more cautious performance evaluations, as thermal and mechanical response of a plate with certain geometry may not be representative for a plate with a different geometry. To evaluate the performance of the plates for various geometric parameters, parametric sensitives studies have been employed. One of the important geometric parameters may have potential effects on the performance is the plate curvature. In this study, curved-plates were parametrically simulated to investigate if this geometric parameter has any effects on overall performance, In particular, radius of curvatures of the plates were varied between the bounding values, and the plates were simulated for comparable irradiation histories. The resulted temperature, deformation, stress-strain results were comparatively evaluated. The results have indicated that preferential deformations occur. This consequently caused shifting of plate centerline on curved plates. The magnitude of centerline shifts increased with increasing plate curvatures.

42 ENGINEERING↗

DOE OSTI2019

Effects of Plate Curvature on Thermo-Mechanical Performance of U-10Mo Monolithic Fuel Plates

42 ENGINEERING↗

DOE OSTI2019

Re-Evaluating the Current Safety Goals

The Safety Goals adopted by the US NRC consist of two qualitative safety goals (QSGs) backed up by two quantitative health objectives (QHOs). The QHOs establish risk limits for severe accidents in terms of their radiological consequences, in particular, the average individual health risks of early fatality and latent cancers from radiation exposure of members of the public living in the vicinity of a nuclear power plant. This paper is devoted to a re-examination of the coverage of the current safety goals in evaluating the actual risk posed by nuclear power plant operation. Experience of recent Level 3 PRAs suggests that given a high likelihood of evacuation of the close-in population before any release occurs, the current QHOs are satisfied by large margins. However, the experience of an actual severe accident at Fukushima showed that actual human health effects from released radiation were not the dominant consequences; there were no early fatalities and no measurable increases expected in cancer rates above the baseline rates in the Japanese population. Hence, regardless of accident probability, Fukushima-type accidents would satisfy the NRC’s health-related safety goals. However, there were very significant societal costs, in that large numbers of people were relocated for long periods. We argue that, in addition to the risks addressed in current safety goals, societal risk should also be evaluated. The paper discusses specific possibilities for a goal and an associated quantitative objective.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

A PROVEN APPROACH FOR EFFECTIVE COMPUTER SECURITY SELF-ASSESSMENTS AT NUCLEAR FACILITIES

A proven method for conducting cybersecurity self-assessments at nuclear power plants is now available for international use. This method was originally developed by Pacific Northwest National Laboratory, under the sponsorship of the U.S. Nuclear Regulatory Commission (NRC), for use at U.S. nuclear power plants. The “Method,” described in NUREG/CR-6847 “Cyber Security Self-Assessment Method for U.S. Nuclear Power Plants,” was originally a limited release document that was withheld from public disclosure but is now publicly available. The Method provides a systematic, phased, and risk-informed approach to help decision makers and security specialists understand their relative cybersecurity posture. Completed Method assessments may be used to support or validate selection of computer security controls to mitigate cyber threats as well as demonstrate compliance with regulations or statutes enacted by competent authorities. The Method assesses the cybersecurity posture of key systems at a nuclear facility with a focus on protection of design base functions. It considers both physical and digital elements of system vulnerabilities and the resulting potential consequences from exploitation. It is well-suited for addressing blended cyberattacks. A semi-quantitative analytical approach is used in the evaluation of potential vulnerabilities, consequences, and risks and provides a technical basis for the selection of security controls to mitigate cyberattacks. The Method’s application at U.S. nuclear power plants has been very encouraging. The only nuclear plant in the United States that did not have adverse findings during its initial NRC computer security inspection prepared for inspection through the diligent application of this self-assessment method and the implementation of recommendations that came out of that self-assessment. Nuclear facilities around the world might find application of the Method extremely helpful for making cost-effective, risk-based decisions regarding computer security and for preparing to pass computer security inspections by their competent authorities.

Cyber security, Nuclear security, Self-assessment↗

DOE OSTI2018

Hydropower Operations in the Colorado River Basin: Institutional Analysis of Opportunities and Constraints

The Colorado River Basin is facing an unprecedented drought. In ongoing drought management efforts, limited attention has been paid to hydropower generation. While some studies do exist on hydropower, they are quantitative in nature and focus on calculating the reduction in megawatts generated at dams in the Basin with declining water availability. These studies simplify the complex process of hydropower generation; water availability is but one factor that impacts hydropower generation. At a more fundamental level, formal institutional arrangements, that is, laws, policies, and rules create the framework within which dams are operated and hydropower is generated. This paper conducts a comparative institutional analysis of water, environment, and energy laws and policies and changes therein to understand the constraints and opportunities faced by hydropower generation in the Colorado River Basin. To tease out the nuances in how institutional arrangements affect dam operations and hydropower generation, the comparative analysis focuses on the two largest and strategically important dams in the Basin: Hoover and Glen Canyon. This paper uses Elinor Ostrom’s Institutional Analysis and Development Framework to analyze laws and policies at three levels: constitutional-choice, collective-choice, and operational levels. Constitutional-choice level laws and policies apply to the entire Basin, whereas collective-choice level and operational level laws and policies are dam specific. Hoover and Glen Canyon Dams face similar biophysical challenges by the virtue of their location in the same river basin. Yet, despite the similarity in the biophysical setting, the analysis in this study finds that the differences in the applicability of constitutional-choice level laws along with the differences in dam specific collective-choice and operational level institutional arrangements produce a distinct set of constraints for hydropower generation at Hoover and Glen Canyon Dams. Even without a drought, water and environmental laws at both the constitutionalchoice and collective-choice levels as well as power contracts constrain hydropower generation and limit the flexibility with which Glen Canyon Dam can be operated. Water and environmental laws also impose specific water release requirements that, at times, require off-peak power generation at Glen Canyon Dam. On the other hand, even with a drought, Hoover Dam faces limited hydropower generation constraints and can operate flexibly. This is because constitutional-choice level laws and dam-specific collective-choice and operational level laws pose limited constraints for flexible daily operations at Hoover. The result is that Hoover Dam can generate hydropower at the same level as it did three decades ago and operate flexibly to provide ancillary services and peaking generation. While water and environmental laws and policies pose constraints for hydropower generation, the analysis in this study further finds that specific historic provisions within energyrelated institutional arrangements and recent changes within power contracts have maintained and even enhanced the value of hydropower to power customers. Historic institutional provisions ensure that hydropower is sold ‘at cost’ making this resource economically competitive with wholesale electricity market rates. Recent power contract modifications have resulted in the amendment of an older resale prohibition clause to expand the flexibility available to power customers in using their capacity and/or energy allocation in RTOs, ISOs, and bulk power markets. This amendment has opened up an opportunity for customers, especially Hoover power customers, to use flexible generation and ancillary services in a market environment. In addition, the extension of power contract duration to the legally maximum term has enhanced the reliability and stability of this resource for customers. In the Colorado River Basin, despite the enduring economic responsibility of power customers—where laws require customers to pay for a large portion of construction and O&M costs whether or not they actually receive hydropower— the persistent threat of a drought-induced water shortage, and constraints imposed by water and environmental laws and policies, power customers continue to invest in this resource as energy-related institutional arrangements and power contract provisions protect the reasons why they value hydropower. Lastly, the analysis in this study finds that the consequences of changes in hydropower generation for energy users, irrigators, and environmental programs in the Basin depend on how specific institutional arrangements tie electricity revenues to irrigation aid and environmental programs, and how the power contracts themselves are set up. Collective-choice level institutional arrangements create a higher level of financial dependency of irrigation aid and environmental programs on electricity revenues in the Upper Basin—the legal subdivision of Colorado River where Glen Canyon Dam is located—compared to the Lower Basin—the legal subdivision of Colorado River where Hoover Dam is located. Therefore, changes in hydropower generation or the way its revenue is collected and used will have far reacting detrimental consequences for the Upper Basin. Likewise, differences in the nature of power contracts for Glen Canyon and Hoover Dams also creates differences in the financial impact incurred by energy users when there is a reduction in hydropower generation. While this study identifies the types of impacts on resource users as a result of specific institutional arrangements, the calculation of extent of impact warrants further attention. Hydropower in the United States is in a unique position today. The strategic importance of this resource for the nation’s electricity sector is rapidly growing even as its contribution to overall electricity generation remains fairly small. This strategic importance, however, is built hydropower’s ability to operate flexibly in order to support the integration of intermittent renewable generating sources and the expansion of electricity markets. As this study shows, such flexibility may not be available at certain plants not due to the lack of water availability but because of institutional constraints. Institutional arrangements may also require dam operators to first consider high priority water uses (such as irrigation or environmental needs), which in turn may limit the ability to generate hydropower when it is most valuable or useful. Engineering and quantitative models, such as production cost models, recognize policy constraints for hydropower operations but often inadequately capture or assume away such constraints in the models. A failure to account for policy constraints in these models runs the risk of inaccurate representation of the operational flexibility and capacity available at specific hydropower plants, which can result in over/underestimation of hydropower’s ability to support the integration of variable renewable resources and address grid reliability concerns. Against this background, this paper and the analysis herein serves as an example of how we can systematically identify institutional constraints (and opportunities) that influence the flexibility in not only generating electricity at specific dams but also using this hydropower once it is generated.

13 HYDRO ENERGY↗