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DOE OSTI2016

Integration of Human Reliability Analysis Models into the Simulation-Based Framework for the Risk-Informed Safety Margin Characterization Toolkit

This report presents an application of a computation-based human reliability analysis (HRA) framework called the Human Unimodel for Nuclear Technology to Enhance Reliability (HUNTER). HUNTER has been developed not as a standalone HRA method but rather as framework that ties together different HRA methods to model dynamic risk of human activities as part of an overall probabilistic risk assessment (PRA). While we have adopted particular methods to build an initial model, the HUNTER framework is meant to be intrinsically flexible to new pieces that achieve particular modeling goals. In the present report, the HUNTER implementation has the following goals: • Integration with a high fidelity thermal-hydraulic model capable of modeling nuclear power plant behaviors and transients • Consideration of a PRA context • Incorporation of a solid psychological basis for operator performance • Demonstration of a functional dynamic model of a plant upset condition and appropriate operator response This report outlines these efforts and presents the case study of a station blackout scenario to demonstrate the various modules developed to date under the HUNTER research umbrella.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2016

Preliminary analysis of core heat-up for ATFs under accident conditions

The accident-tolerant fuel (ATF) system is aim at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the conventional UO{sub 2}-Zr fuel system when the reactor is in normal operation. In this paper, the improved performance of UO{sub 2}-SiC, UO{sub 2}-FeCrAl fuel systems, and of fully ceramic micro-encapsulated (FCM) fuels: FCM-SiC and FCM-FeCrAl, on the core heat-up progress in a pressurized water reactor (PWR) is analyzed by using the best-estimate system code RELAP5 and severe accident analysis program. For this purpose two accident scenarios are selected: a large break LOCA as a design base accident and an extended station blackout (SBO) severe accident. For the LBLOCA with just low-pressure safety injection, larger volumetric heat capacity and thermal conductivity of ATF materials contribute to a larger safety margin in terms of the peak cladding temperature compared with UO{sub 2}-Zr. Thanks to higher melting temperatures, FCM-SiC is failed by melting almost 1200 s later than UO{sub 2}-Zr. The accumulated production of hydrogen for ATF materials is around 10 times smaller than that for UO{sub 2}-Zr. However, high temperature oxidation of SiC cladding generates carbon monoxide which is combustible, while the melting point of FeCrAl is not high enough. In the SBO scenario with mitigation measure, the advantage of ATF is more evident. If the mitigation measure could not be taken, core heatup is inevitable for all types of fuel system, but the temperature rising of ATF core is much slower. In general, improved high-temperature properties and resistance to oxidation play important roles for ATF cores.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Preliminary modeling of accident tolerant fuel concepts under accident conditions

In this paper, we present simulations of two proposed accident tolerant fuel systems for light-water reactors: U{sub 3}Si{sub 2} fuel with Zircaloy-4 cladding, and UO{sub 2} fuel with FeCrAl cladding. The simulations investigate the fuel performance response of the proposed ATF systems under Loss of Coolant (LOCA) and Station Blackout (SBO) conditions using the BISON code. It was observed that U{sub 3}Si{sub 2}/Zircaloy4 and UO{sub 2}/FeCrAl have superior performance during a 90 s LOCA than traditional UO{sub 2}/Zircaloy4 fuel rods. This is based upon reduced fuel centerline temperatures and lower cladding strains experienced during the LOCA. For the postulated SBO analyzed in this work, it was found that the centerline temperature and cladding strains remained well below any values of concern. This is attributed to the shortness of the rod studied and how the coolant channel model treats the inlet as the bottom of the rod. An analysis of a full-length fuel rod is required to obtain improved qualitative comparisons between the 3 systems under SBO conditions. Sensitivity analyses have been performed using Sandia National Laboratories' DAKOTA software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). Early results indicate that each concept has significant advantages as well as areas of concern. Further work is required prior to formulating the proposition report for the Advanced Fuels Campaign.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Knowledges and Insights Obtained from the Fukushima Daiichi Accident

It was very hard to estimate how the accident had progressed at each unit of Fukushima Daiichi NPS at the early stage of post-accident analytical activities, because the knowledge relevant to BWR severe accident phenomena and progression was insufficient and the collected data was very limited since complete station blackout (loss of AC and DC power) was occurred at Fukushima Daiichi NPS. However, because of the progression of the measured data analysis, system behavior or response of reactors before the core melt has already been clarified. System behaviors after the core melt have been clarified, as well. Measured data and circumstantial evidence clearly indicated that the fuel temperature went up to melting point and the RPV was damaged by the molten core, and the Fukushima accident is known as a more serious accident than TMI-2 because of including ex-vessel phenomena. But, the information regarding detailed accident progression (for example, amount of melted fuel and timing of RPV breach, etc.) is still unclear and we are struggling to solve such a detailed accident progression. At present, there are two challenges in front of us. One of them is to improve nuclear reactor safety by applying the insights from the Fukushima accident and the other is to solve unclear issues before conducting the first RPV internal investigation. Tokyo Electric Power Company is willing to stand up against these two challenges as the company concerned and appreciate supports of the world nuclear community. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Severe Accident Research Directions after Fukushima

The accidents at the Fukushima Daiichi nuclear power station in March of 2011 renewed interest globally in the nature, causes, and health and economic consequences of severe accidents involving core damage and containment impairment. As a result of these accidents, research has been conducted into the modeling and phenomenological understanding of the accidents at Fukushima and the response and effectiveness of mitigating measures such as emergency water injection, primary system depressurization and containment venting actions. An important example of this is the jointly conducted OECD-NEA organized Benchmark of the Fukushima Accidents (BSAF project) where accident reconstruction analyses using computer codes such as MELCOR and MAAP are ongoing. Even prior to the Fukushima accidents, the Sandia/NRC sponsored State of Art Reactor Consequence Analyses (SOARCA) foreshadowed the accidents in Japan where long-term and short-term station blackout accidents in boiling water reactors with Mark-1 containment were analyzed showing striking similarities to the accidents at Fukushima. These analyses also highlighted important differences in our prior understanding, in particular the importance of suppression pool thermal stratification and the extraordinary operation of the Fukushima Unit-2 reactor core isolation cooling (RCIC) system, prompting addition research into both of these areas. Additionally, the focus of the world's major computer codes on modeling the Fukushima accidents in the BSAF effort, led to several 'Crosswalk' comparisons between code predictions, highlighting subtle but important modeling differences between the codes and identifying areas where modeling improvements can be made. Collectively, these research areas are aimed at improving our understanding of severe accident progression and real-world response of safety related systems and components, and should lead to improved accident response and mitigation capabilities in the future. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Experimental validation on the operational and cooling performance of the passive safety system with Atlas

After the Fukushima accident, passive safety systems are considered as the most promising alternatives to reinforce the safety and reliability of an ultimate heat removal system without any operator actions. In Korea, a passive auxiliary feedwater system (PAFS) was developed and adopted to a GEN-III+ nuclear power plant, APR+ to completely replace the conventional auxiliary feedwater system. The integral effect tests were performed to validate the cooling and operational performance of PAFS, coupled with the reactor coolant systems using the ATLAS facility. The representative design basis accidents (DBAs) such as a feedwater line break, a steam line break, and a steam generator tube rupture transients were experimentally simulated in this study. In addition, an integral effect test was performed for simulating a station blackout transient as a limiting beyond DBA. From the integral effect test results, it could be concluded that the APR+ has the capability of coping with the hypothetical DBA and bDBA by adopting PAFS and proper set-points of its operation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

The MELCOR Analysis of Chinshan Nuclear Power Plant Spent Fuel Pool for Fukushima-like Accident

The safety analysis of the nuclear power plant (NPP) is very important work in the NPP safety. Especially after the Fukushima NPP accident occurred, the importance of NPP safety analysis has been raised and there is more concern for the safety of NPPs in Taiwan. Because the earthquake and tsunami occurred, the cooling system of spent fuel pool failed and the safety issue of spent fuel pool generated in Japan's Fukushima NPP. Chinshan NPP is the first NPP in Taiwan which is BWR/4 plant and OLTP (Original Licensed Thermal Power) for each unit is 1775 MWt. Chinshan NPP finished SPU (stretch power uprate) and the operating power is 103.66% of the OLTP, which is 1840 MWt now. After Fukushima NPP accident occurred, in order to concern the safety of Chinshan NPP spent fuel pool, we performed the safety analysis of spent fuel pool by using MELCOR/SNAP which also assumed the cooling system of spent fuel pool failed. The geometry of Chinshan NPP spent fuel pool is 12.17 m x 7.87 m x 11.61 m and the initial condition is 60 deg. (water temperature) / 1.013 x 10{sup 5} Pa. And, the total power of the fuels is roughly 8.9 MWt initially. MELCOR is developed by Sandia National Laboratories. MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor NPPs, including containment response and leakage of activity to the environment. A broad spectrum of severe accident phenomena in both BWRs and PWRs is treated in MELCOR in a unified framework. These include thermalhydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, and combustion; fission product release and transport behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. According to Dr. Carbajo's paper, MELCOR was used to perform the study of spent fuel pool of Fukushima Daiichi Unit 4. Niina E. Koenoenen used MELCOR to establish the spent fuel pool model of a Nordic BWR. The loss-of-pool-cooling accidents was simulated and analyzed by this model. The above studies indicate that MELCOR is capable of handling the simulation of the spent fuel pool. An increasing number of researchers are using MELCOR code to analyze test facilities and nuclear power plants. Jun Wang et al. used MELCOR to perform core degradation simulation and MELCOR results were compared with CORA experimental data. Through this work, they reviewed the performance of MELCOR COR package in detail. Tae Woon Kim et al. developed MELCOR model of APR1400 (an advanced pressurized water reactor). They used this model to perform the LBLOCA transient simulation and sensitivity study. Longze Li et al. established the MELCOR model of CPR1000 (a Chinese pressurized reactor 1000-MW power plant). The station blackout (SBO) with failure of the steam generator (SG) safety relief valve (SRV) transient was simulated and analyzed by this model. SNAP is a graphic user interface program that processes the inputs and outputs of MELCOR. In addition, there is an animation function in SNAP which can present the animation of analysis results. There are two main steps in this research. The first step was the establishment of Chinshan NPP spent fuel pool MELCOR/SNAP model. And the transient analysis of Chinshan NPP spent fuel pool MELCOR/SNAP model under the cooling system failure condition was performed. The next step was the comparison among the MELCOR, TRACE and CFD data. TRACE and CFD data were from INER reports. In addition, the animation model of Chinshan NPP spent fuel pool was presented by using the animation function of SNAP with MELCOR/SNAP analysis results. This study has developed MELCOR/SNAP model of Chinshan NPP spent fuel pool successfully. By using the above model, the safety analysis of the spent fuel pool was performed under the cooling system of spent fuel pool failed condition. The analysis results of MELCOR, TRACE and CFD were similar in this case. It indicated that there was a respectable accuracy in MELCOR/SNAP model. The analysis results depicted that the uncovered of the fuels occurred at 2.6 day and the metal-water reaction of fuels occurred roughly at 3.7 day after the cooling system failed. The above results indicated that the failure of cladding occurred after 3.7 day. This study's results can help to evaluate the safety issue of Chinshan NPP spent fuel pool. In addition, the cooling system and some mitigation measures of spent fuel pool is not simulated in the above model. However, we will do it in the future. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Analysis of ELAP and mitigation strategies in Maanshan PWR with PCTRAN

On March 11, 2011, the Fukushima Daiichi Nuclear Power Plant, owned by Tokyo Electric Power Company, encountered a beyond design basis earthquake/tsunami and the flood crippled the on-site emergency power and electric transmission systems. The ELAP (Extended Loss of Alternating Current Power) and LUHS (Loss of Ultimate Heat Sink) events severely damaged the core cooling systems and the safety functions of the containment. As a result, three reactor cores were damaged. Also, the cooling capability of nuclear spent fuel pools was lost due to loss of AC and DC power. Existing Station Blackout (SBO) analyses have demonstrated that even with leakage from Reactor Coolant Pump (RCP) seals, the core could maintain coolable only when heat transfer exists between primary loop and secondary loop. Depend on the availability of DC power in different plant designs, the SBO coping times were shown to be 4 to 16 hours. According to INPO event report, it is necessary to analyze the consequences if current coping times are extended with increased core water inventory loss from pump seals. In an ELAP scenario, all active system including Emergency Core Cooling System (ECCS) and motor driven AFW were out of power supply, results in degradation of core cooling capability. Moreover, the fuel exposed risk would increase due to continuously Reactor Coolant System (RCS) inventory loss through pump seals. Therefore, mitigation strategy must be applied prior to core uncovering to prevent its damage and assure plant safety. This research utilized a fast running NPP safety simulation code, PCTRAN, to assess reactor performances in the situations of ELAP of Maanshan PWR. Also, the sensitivity study of injection flow rates and effectiveness of mitigation strategy were demonstrated. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Experimental and CFD Analysis of Condensation with Non-Condensable Gas in Multi-Tubes

After the Fukushima accident, the importance of safety system that relies on passive cooling for decay heat removal has been reemphasized. The passive containment cooling system (PCCS) is the system for cooling down a containment and lowering a containment pressure even under the station blackout. In investigating a PCCS, steam condensation in the presence of non-condensable gas is the dominant heat and mass transfer phenomenon. The non-condensable gases degrade the heat removal performance through the PCCS heat exchangers (HXs) because they are accumulated near the cooling surface and deteriorate the approach of steam by diffusion to the surface. Therefore, in order to understand steam condensation phenomenon in the PCCS, we conducted multi-tube condensation experiment with air. And we also analyzed the results by CFD simulation for the CFD-based approach to condensation phenomena prediction. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Pressure Drop-Flow Rate Curves for Single Phase Steam in Steam Generator U-tubes During Severe Accidents

Thermally induced steam generator tube rupture (TI-SGTR) can occur in pressurized water reactors (PWRs) with U-tube steam generators (SGs) during high pressure severe accident sequences resulting in containment bypass. Robust natural circulation patterns including in-vessel natural circulation and countercurrent flow to and from the SGs of superheated steam and other hot gases heats cooler structures in the reactor coolant system (RCS). The buoyancy driven flows return cooler and denser gases back to the core slowing the progression of some core degradation phenomena and failure of the lower head. As much as 50 percent of the decay energy and energy produced from core degradation oxidation reactions can be transported from the core to SGs resulting in the heat up and thermal creep of RCS pressure boundary components including hot leg nozzles, pressurizer surge line, and the SG tubes, many of which contain flaws. The creep failure of an individual tube is dependent on the transitory heating of the tube, a function of the mass flow rate through the tube and local temperature history of steam entering the tube. Buoyancy driven natural circulation flow rates are determined by changes in gravitational head, a combination of changes in fluid density due to heat transfer and change in elevation, and irreversible pressure drops resulting primarily from friction loss in the flow paths and other form losses. The nonuniform flow distribution in U-tubes of varying lengths and heights during single-phase liquid and two-phase natural circulation has been extensively studied in scaled test facilities and verified numerically in previous works. We hypothesize the countercurrent natural circulation of single-phase superheated steam in U-tube SGs is governed by the same characteristic pressure drop-flow rates curves of single-phase liquid and two-phase natural circulations in U-tubes connected to common plena derived in. Secondly, we propose flow distribution is a function of U-tube geometry - height, length, hydraulic diameter, etc., and the larger SGs of Combustion Engineering (CE) style 2 x 4 PWRs, which are the focus of this study, will show greater nonuniformity in tube by tube flow rates and ratio of forward flowing tubes to reverse flowing tubes. This work derives characteristic pressure drop-flow rate curves for single-phase superheated steam in CE-type SGs during a high pressure station blackout (SBO) sequence with failed auxiliary feedwater (AFW) and dry secondary side which is closely related to the TI-SGTR event. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Estimation of Core Recovery Time with TRACE

Following the severe accident in Fukushima Daiichi Nuclear Power Plant in Japan, safety related concerns about the zirconium based claddings used in current light water reactors have increased. High corrosion resistance and low neutron absorption cross section make zirconium a suitable cladding material for normal nuclear plant operational conditions. However, for nuclear accident scenarios when safety functions of the plant are lost, Zr claddings undergo exothermic reaction with steam and exacerbate the accident conditions. The exothermic reaction increases the temperature of the cladding and accelerates core melting in severe accident progression. And the generation of explosive hydrogen gas accumulates and challenges the structural integrity. The accident tolerant fuel (ATF) concept is the fuel and cladding system that gives a longer time margin, compared to standard Zr claddings, to cope with the accident when core cooling systems are lost in a nuclear power plant. The ATF cladding should have slower oxidation kinetics and less hydrogen generation than Zr to alleviate consequences of cladding interaction with steam at high temperatures. The maximum time available to mitigate progression of severe accident is an important measure to evaluate benefits of ATF claddings. For this work, the maximum time available to mitigate progression of severe accident is selected as the time at which cladding melting reaction starts. Melting reaction is selected as an upper limit because of the degradation of coolable geometry. Maintaining the coolable geometry is a regulation criteria for both loss of coolant (LOCA) analyses and outside the LOCA analyses. Once the melting of the cladding starts, it exacerbates the deformation of the fuel assembly and hottest rods within the core cannot be cooled by water reflood and results in fuel melting progression. The aim of this work is to evaluate time margin gain of ATF cladding in a severe accident scenario of LOCA without safety injections and station blackout considering oxidation kinetics only. Also it is assumed that safety injections are recovered before melting of the claddings start and the sensitivity of safety injection actuation time on core recovery is analyzed. The core recovery time is the time when makeup water from safety injections quench the fuel rods and halt the progression of the severe accident. This time is estimated by initiating safety injections at different times and observing the hot rod behavior which represents 1/5. of the core in this work. It is considered that core is not recoverable when 20% of the core is damaged.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Optimizing power system restoration with damaged communications

Utility procedures for power system blackstart and restoration typically assume that energization decisions can be reliably communicated across the grid. In reality, the communications and control network would likely also be affected in power outages, such as those caused by extreme weather events or cyber-attacks. This paper studies the effect of damage to the power system communications and control infrastructure on restoration operations following a blackout. We model the communications infrastructure as a graph, overlaying the power grid, and imposing the requirement that every energized element in the power grid be observable from a control center. We expand on a specialized branch-and-bound algorithm from the literature to optimize the restoration process and devise an initialization heuristic and a rounding heuristic to improve solution speed. We perform numerical experiments on synthetic systems for Illinois and Texas with outages based on a solar flare or hurricane. We compare the results of our specialized branch-and-bound algorithm to the results from (i) the initialization heuristic alone, (ii) a variation of this heuristic that we use as a baseline, and (iii) the restoration optimization for the power system without communications constraints. Here, we find that damage to the communications infrastructure significantly increases the time required to re-energize the grid. Moreover, by simultaneously optimizing communications repairs and grid energization decisions, we are able to re-energize the grid significantly faster than if communications repairs and energization decisions were made independently or with partial coordination, motivating improvements to current industry practice.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

Chapter 32 - Power Grid Resilience

The new energy paradigm is altering the current power grid trend from synchronous generator reliant system toward power-electronics-based distributed energy resources (DERs). The resilient operation of modern power grid is highly dependent upon cyber and physical reliable operation of DERs. Power grid resilience broadly refers to the ability of the grid to be robust to eventualities and singularities that may disrupt the continuity of reliable power flow. This could involve resilience related to the physical system but more recently, the focus on cyber resilience has been evolving rapidly especially given the burgeoning growth of distributed generation and power-electronics-based DERs under smart grid and given that such threats to reliable operation do not have to evolve localized to where the physical asset is. Commonly, in power grids dominated by DERS, control and energy management are structured at a multitime scale multilayer fashion: (i) primary control layer at microseconds time scale, (ii) secondary control layer at millisecond time scale, and (iii) tertiary control layer at seconds to minutes time scale. The cyber-related issues that may affect the power grid resilience can be introduced in through primary, secondary, and tertiary control layers. The failure of each of these control layers may impact the reliable and resilient operation of the overall network and cause widespread failures which leads to unintended blackouts. As such, this chapter focuses on cyber-related resilience issues in primary control layer, secondary control layer, tertiary control layer, wide area/utility control, and anomaly detection and resilient communication.

anomaly detection↗
DOE OSTI2018

Phenomena identification and ranking table study for thermal hydraulics for Advanced High Temperature Reactor

The Advanced High-Temperature Reactor (AHTR) is a conceptual design of a Fluoride salt-cooled High-temperature Reactor (FHR) utilizing 7 LiF-BeF 2 (66-34 mol%) as its primary coolant. To identify key phenomena that impose potential challenges on thermal hydraulics modeling and simulation of such a reactor system, a thermal hydraulics phenomena identification and ranking table (TH-PIRT) study was performed for the AHTR in a Department of Energy Nuclear Energy University Program (NEUP) integrated research project led by Georgia Institute of Technology. A panel of experts from regulators, industries, national laboratories, and academia was assembled for the study. In this paper, the TH-PIRTs identified by the panel for two events, namely, station blackout and simultaneous withdrawal of all control rods, are summarized and discussed in detail. In addition, the key phenomena that warrant further study and research for AHTR analysis are identified to support the validation of thermal hydraulics system-level analysis codes and computational fluid dynamics simulation tools, as well as future FHR reactor licensing.

42 ENGINEERING↗
DOE OSTI2025

A method for assessing economic, environmental, and reliability tradeoffs of interregional transmission connecting ERCOT (the Texas grid) to the eastern and western grids

Reliable development of the power grid is an evolving concern for humanity due to extreme weather that frequently threatens power sector infrastructure. The state of Texas is a uniquely structured testbed for grid planners to study when looking for solutions to development, innovation, and overcoming such challenges. Because of its size and islanded structure, Texas is small enough to model, but big enough to matter. Texas is a global leader in energy production, energy consumption, and maintains an unusually diverse fuel mix. In addition, the state has experienced winter freezes, heat waves, wind storms, droughts and floods that have threatened power sector infrastructure or caused recent blackouts and calls for demand side conservation. One of the most devastating of these events was the North American winter storm, dubbed “Winter Storm Uri” by the Weather Channel, that froze the region in February 2021 and led to an extended power outage event that put the majority of Texan residents in darkness for days. While preparing to avoid such outage events in the future, various tools have been proposed to improve grid reliability, including energy efficiency, demand response, and distributed energy resources. An additional option would be to develop interregional transmission that connects the Texas grid to other national grids. To assess the merits of this idea, we developed a novel, universally-applicable and internationally-relevant framework to study how the Texas grid would evolve alongside access to various interregional ties. This method allows us to stress the synthetic grid structure and analyze how it would respond to the shock of a simulated winter storm event. Our method leverages open-source modeling tools, such as PowerGenome, pyGRETA, and GenX to synthesize unique zonal grid data, construct a consolidated network of model regions, and simulate different developmental pathways of capacity expansion and operational dispatch. We demonstrate our method with an analysis connecting the Electric Reliability Council of Texas (ERCOT), the grid that serves most of Texas, the Western Electricity Coordinating Council (WECC), the grid that serves the western half of the contiguous U.S., and the Eastern Interconnect, the grid that serves the eastern half of the contiguous U.S. Our results indicate that the cost-optimal capacity of interregional transmission connecting the ERCOT grid to other grids lies between 9–13 GW assuming baseline conditions. Building this amount of connecting capacity in one or multiple directions lowers the costs and emissions of development and operation by up to $16 billion and 257 million metric tonnes (MMT) respectively. Additionally, our results show that the interregional connections between ERCOT and other national grids reduce the amount of total load shed required through mild winter storm events. However, our results also show that there is a threshold of very extreme winter storm conditions, spanning multiple service areas, above which the connections exacerbate resource adequacy problems. Therefore, the results indicate that the connections need to be carefully planned alongside the rest of the grid infrastructure to avoid over-reliance on specific resources or technology options.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2026

Consumer safety-oriented scheduling of rotating power outages during heat waves

Extreme heat events have widespread effects on power systems, reducing available generation capacity, limiting transmission capabilities, and causing unusual demand patterns on the consumer side. As these combined effects expose bulk transmission systems to potential large-scale blackouts, utilities may be required to schedule and apply rotating outages, by temporarily and alternately disconnecting distribution substations to reduce overload. However, utilities lack mechanisms to inform these events, exacerbating the negative effects of heat waves on affected communities. This paper introduces a novel framework for scheduling rotating outages during heat waves while considering impacts on consumers’ safety. Instead of random sequential load shedding, we propose a methodology to rotate power outages considering a metric that quantifies the indoor overheating risk of groups of consumers during a power outage. The overheating risk is derived from a detailed building simulation using CityBES, where the buildings are modeled based on available data—use type, year built, floor area, number of stories, location—while presence of air conditioning and occupancy are calibrated from smart meter data. Based on the metric, an algorithm to schedule the rotating outages is applied to prioritize feeders for disconnection at each hour according to their overheating risk to meet a utility load reduction target. Applied to two substations and seven feeders in the Portland General Electric territory, the results show that this approach effectively leads to the lowest overheating risk during the resulting outage schedules, with an average 10.1% lower overheating compared to uninformed schedules.

Building thermal simulation↗
DOE OSTI2023

Mechanistic understanding of aging behaviors of critical-material-free Li 4 Ti 5 O 12 //LiNi 0.9 Mn 0.1 O 2 cells with fluorinated carbonate-based electrolytes for safe energy storage with ultra-long life span

Behind-the-meter storage (BTMS) systems - a viable method to minimize potential risk of blackout events and stabilize the grid - require a different type of cost-effective energy storage with excellent safety, ultra-long (>20 years) cycle life and reasonable energy density compared that of electric vehicles. To increase the energy density and reduce the cost of a long-term cyclable lithium-titanate-based cell, it is required to employ a critical-material-free high voltage cathode and an electrolyte with good electrochemical and transport properties. In this report the long-term electrochemical performance and behaviors of selected critical-material-free Li 4 Ti 5 O 12 (LTO)//LiNi 0.9 Mn 0.1 O 2 (LNMO) full cells for BTMS applications are evaluated and analyzed in the optimized voltage range of 1.4-2.7 V at 45 degrees C with different fluorinated carbonate-based electrolytes. The fluoroethylene carbonate (FEC)-based electrolyte cell shows the highest capacity retention of 57.9% and Coulombic efficiency (CE) of 99.96% after 1000 cycles, potentially attributed to a dense, homogenous and less resistive LiF-rich solid-electrolyte interphase (SEI) layer formed on the surface of LTO that may mitigate electrolyte decomposition and maintain relatively low cell impedance during cycling. The 3,3,3-fluoroethylmethyl carbonate (F-EMC)-based electrolyte cell, however, presents the worst performance with lower capacity and a sharp decrease of CE, due to unstable and non-uniform SEI formation and continuous oxidative electrolyte decomposition. This mechanistic understanding of cell aging behaviors and failure mechanisms with detailed analysis of surface chemistry and electrode morphology can guide design of new electrode chemistries and electrolyte formulations for the development of BTMS batteries.

25 ENERGY STORAGE↗
DOE OSTI2023

Off-policy deep reinforcement learning with automatic entropy adjustment for adaptive online grid emergency control

Electric overloading conditions and contingencies put modern power systems at risk of voltage collapse and blackouts. Load shedding is crucial to maintain voltage stability for grid emergency control. However, the rule- or model-based schemes rely on accurate dynamic system models and face considerable challenges in adapting to various operating conditions and uncertain event occurrences. Here, to address these issues, this paper proposes a novel deep reinforcement learning (DRL)-based voltage stability control algorithm with automatic entropy adjustment (AEA) for grid emergency control. Various dynamic network components for complex system operations are modeled to construct the DRL environment. An off-policy soft actor-critic architecture is developed to maximize the expected reward and policy entropy simultaneously. The AEA mechanism is proposed to facilitate the policy maximum entropy procedure, and the proposed method can automatically provide effective discrete and continuous actions against various fault scenarios. Our approach accomplishes high sampling efficiency, scalability, and auto-adaptivity of the control policies under high uncertainties. Comparative studies with the existing DRL-based control methods in IEEE benchmarks indicate salient performance improvement of the proposed method for dynamic system emergency control.

24 POWER TRANSMISSION AND DISTRIBUTION↗