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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

DYNAMIC PRA OF A MULTI-UNIT PLANT

Dynamic Probabilistic Risk Analysis (PRA) methods couple stochastic methods (e.g., RAVEN, ADAPT, ADS, MCDET) with safety analysis codes (RELAP5-3D, MELCOR, MAAP) to determine risk associated to complex systems such as nuclear plants. Compared to classical PRA methods, which are based on static logic structures (e.g., Event-Trees, Fault-Trees), they can evaluate with higher resolution the safety impact of timing and sequencing of events on the accident progression. Recently, special attention has been given to nuclear plants which consist of multiple units and, in particular, on the safety impact of system dependencies, shared systems and common resources on core damage frequencies. In the literature, while classical PRA methods have been employed to model multi-unit plants, Dynamic PRA methods have never been applied to analyze a full multi-unit model. This paper presents a PRA analysis of a multi-unit plant using Dynamic PRA methods. We employ RAVEN as stochastic method coupled with RELAP5-3D. The plant under consideration consists of the three units and their associated spent fuel pools (SFPs). The studied initiating event is a seismic induced station blackout event. We will describe in detail how the multi-unit plant has been constructed and, in particular, how unit dependencies and shared resources are modeled.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2017

Convolution Correction Factor Adjustments on Static PRA Models for Event Assessment

Current probabilistic risk assessment (PRA) models will contain cut sets that have time-constrained basic events such as switched components, recovery of failed components, and failures in time. For these cut sets that have multiple time-constrained events that interact (e.g., offsite power recover at the same time as multiple diesel generators fail to run), there is the possibility that the minimal cut set quantified result is too large. Since current PRA models typically assume that the accident scenario starts at “time zero,” the fails-to-run basic events use this starting time as the basis for their respective reliability models. Over time, the standard practice was to recognize this issue and accept it as part of the conservatism built into the event tree/fault tree PRA models. However, now that these same PRA models are being used in a regulatory forum, this conservatism needs to be addressed. Further, this issue exists not only for the baseline PRA results, but for deficiencies that have been observed or those that have the potential of occurrence, whether the deficiency is complete failure or just a degraded condition. We have found that this conservatism becomes very important when evaluating scenarios such as station blackout sequences and other loss of offsite power sequences. This paper will address the theory and implemented process used by the Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) PRA software. Starting with SAPHIRE version 8, the process that was developed is to calculate a “convolution correction factor” based on three distinct calculations. The first calculation is the cut set based frequency calculated via the standard event tree/fault tree PRA model (assuming all run failures start at time equal to zero). The second calculation is to perform the exact time-constrained calculation by integrating the component failure models over the time period of interest (e.g., time to recover, a mission time). For example, this exact calculation for the case of the diesel generators is to convolve their failure models with the offsite power recovery model to obtain the exact frequency. The third calculation is to produce the convolution correction factor by dividing the exact frequency calculation by the standard cut set frequency. This convolution correction factor is then applied to the PRA cut sets in order to remove the time-constrained conservatism. The process used to obtain the convolution correction factor for the baseline PRA is important; however, in this paper we also describe what happens when there is degradation of a component or a potential deficiency. These issues become very important in order to ensure that the convolution correction factor is re-calculated correctly. If the correction factor does not get re-calculated and the same baseline convolution correction factor is used, the final analysis result could be over conservative or under conservative. This paper will discuss how these issues are addressed and provide examples on how SAPHIRE automatically makes the necessary adjustments. MathCad verification results can be computed using the provided equations and input data.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2017

Unleashing the Power of Wide-Area Data Analytics for Enhanced Grid Visibility

We report that the operation of modern electric transmission systems relies increasingly on the use of advanced sensing and synchronized-measurement technologies. These technologies collect and collate a wide variety of data from disperse locations to monitor the health of network, obtain fast and accurate diagnosis of systemic changes, and detect signs of anomalies and instabilities at an early stage. In view of the far-reaching blackout risk from potential damages to critical transmission grid assets (due to severe disturbances, physical attacks, and extreme weather events), stealthy activities that compromise the operation of wide-area sensing, communication, and control systems could cripple the entire cyber-physical smart-grid infrastructure. To this end, achieving robust event identification, ensuring bulk system observability, and developing actionable intelligence (by fostering data-driven decision making under imminent threat of complex, multifaceted attacks and disruptions) remain key challenges for sustaining a resilient power grid operation.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2018

Development of LWR Fuels with Enhanced Accident Tolerance ATF Feasibility Analysis Report Deliverable for the Westinghouse Accident Tolerant Fuel Program

The Westinghouse Electric Company LLC (Westinghouse) accident tolerant fuel (ATF) program utilizes chromium (Cr) coated zirconium alloy cladding with uranium silicide (U 3 Si 2 ) high density/high thermal conductivity fuel for its lead test rod (LTR) program with irradiation beginning in 2019. The lead test assembly (LTA) program will use both SiC/SiC composites and Cr coated zirconium alloy claddings with the high density/high thermal conductivity U 3 Si 2 pellet which will begin in 2022. Over the past several years, Westinghouse has tested the Cr coated zirconium (Zr) and silicon carbide (SiC) claddings in autoclaves and in the Massachusetts Institute of Technology (MIT) reactor and the U 3 Si 2 pellets have been tested in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). High temperature tests at the state-of-the-art Westinghouse facilities in Churchill, PA have been carried out to determine the time and temperature limits for the SiC and Cr coated zirconium claddings. Fuel rod and assembly design in preparation for the LTR and LTA programs is underway as well as licensing efforts with the Nuclear Regulatory Commission (NRC). Finally, accident analyses coupled with economic evaluations for both operating savings as well as fuel savings have been initiated. Cr coated zirconium alloy claddings showed low corrosion rates in both autoclave and in-reactor tests and will be fabricated using commercial processes for the first LTRs for insertion in 2019. Current autoclave results for SiC composite claddings indicate that a corrosion rate of fewer than 2 micrometers per year which meets corrosion requirements may be achieved under normal operating conditions. Modular Accident Analysis Program 5 (MAAP5) calculations indicate that solid fission products can be contained within SiC cladding for up to two hours longer than current Zr based cladding in a station blackout scenario. This additional time is due to the much higher temperature capability of SiC (~2000°C). These two hours can be used to implement additional response options instituted through the FLEX program by reactor operators. While the Cr coated Zr alloy option offers more modest ATF gains (~200°C higher than bare Zr) in similar situations, the coatings do delay the runaway oxidation encountered by uncoated zirconium cladding. Both ATF cladding options also reduce hydrogen production which dramatically reduces primary system and containment pressure and the risk of fission product release beyond containment in the unlikely event of an accident. The lower pressure in the system allows more time to feed cooling water to the core, resulting in the avoidance of fuel melting. The coping time is extended indefinitely as long as the water flow continues. SiC exhibits the highest accident tolerance of any cladding and significant advances in understanding the design, manufacture and accident behavior of SiC cladding have been made. Department of Energy support is needed to continue the on-going experimental work addressing the interaction between U 3 Si 2 and SiC at temperatures >1200°C, the effect of SiC manufacturing methods and manufacturing variability on physical properties, and the ability to predict the mechanical properties of SiC composites based on the composite design. Additionally, experimental work on methods to rapidly fabricate SiC composite structures with high density and reduce the fabrication price of SiC fibers while maintaining a high level of performance is needed. Minimal (<1%) swelling of U 3 Si 2 and subsequent fission gas release has been demonstrated up to 20 MWd/kgU. Irradiation experiments with U 3 Si 2 fuel in ATR to determine these properties at 40 to 50 MWd/kgU are underway. These experiments will provide the data needed to license fuel rod design codes. Research is continuing on production methods for making U 3 Si 2 from UF 6 without going through the U metal step. Oxidation tests of fuel pellets and powder in steam and synthetic air indicate that U 3 Si 2 has a lower oxidation reaction initiation temperature than UO 2 . Options to reduce the U 3 Si 2 oxidation rate are being explored. A hybrid fuel design using pellets with the weight ratio of 30%/70% U 3 Si 2 /UN is currently undergoing testing in the ATR. This approach is being explored to further increase the positive economics of the U 3 Si 2 fuel. In addition to the work supported by the USDOE, research and testing activities are being carried on in a world-wide effort funded by many countries such as Sweden, United Kingdom, Belgium, Netherlands, Spain, Germany, Japan and France. This work is being facilitated through the Westinghouse led Collaboration for Advanced Research on Accident Tolerant Fuel (CARAT) program. Annual meetings are organized by Westinghouse as a venue for presentation of this work and to provide for the cross-fertilization of ideas among the many outstanding researchers in the ATF area. Since SiC, coated cladding and high density fuel options are not currently used in the nuclear industry, support from the government and industry members is needed to further the significant effort of setting new standards. The same is true for the Nuclear Regulatory Commission (NRC) which must review and approve the commercial use of these new fuels since all current regulations are oriented toward Zr/ UO 2 fuel. There are no showstoppers for accident tolerant fuel technologies when considering supply chain, fabrication, enrichment suppliers, production, and licensing. Due to the cost savings from higher density fuel and the accident tolerance of coated claddings, Exelon has identified potential commercial plants to host the LTRs and LTAs. Planning is underway to implement this schedule.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Design of a Microgrid Transition Controller II: System Recovery Under Abnormal Conditions: Preprint

This paper presents a microgrid transition controller for managing emergency operation when the microgrid experiences a system blackout caused by an internal or external fault. The developed transition controller consists of various application function blocks, including normal operation, emergency operation, and coordination between them. The developed microgrid transition controller is validated by a sample microgrid, and two test cases are investigated: islanded black start and grid-connected black start. The simulation results demonstrate the feasibility and effectiveness of the proposed microgrid transition controller for handling emergency operation and transitioning to normal operation.

14 SOLAR ENERGY↗
DOE OSTI2018

RISK-INFORMED EXTERNAL HAZARDS ANALYSIS BY EEVE TOOLKIT

Introducing risk-informed safety analyses for external events can bring sensible benefits to the nuclear power industry easing the regulatory burden introduced in the aftermath of the Fukushima accident. In the framework of the US-DOE Light-Water Reactor Sustainability program, Risk-Informed Margin Characterization project (LWRS-RISMC), research activities were conducted at INL for developing and applying a toolkit for performing risk-informed external hazards analysis. The “EEVE” (External EVEnts) toolkit, a chain of INL and state-of-the-art codes, was set-up and applied for simulating earthquake-induced transients with internal flooding in a PWR. This was done using methodology set-up for evaluating new data and determining critical areas that would benefit most from advanced analysis methods. The methodology is agnostic and it can be applied to other external hazards than just seismic and flooding (e.g., high winds, intense precipitation, etc.). The first step was the deterministic calculations for evaluating the propagation of a set of earthquakes and for determining the effects on a fire suppression system in an auxiliary building. These deterministic calculations were performed using the LS-DYNA and the OPENSEE structural mechanics codes. Probability failures and event trees of a generic 3-loops Westinghouse PWR were derived from a generic SAPHIRE model. Seismically induced Loss of Offsite Power, Station Blackout, and fire suppression piping failure events were considered. SAPHIRE code allowed a pre-screening of all possible branches of the event tree and a selection of those with the largest Core Damage Frequency increase for further analysis using dynamic PRA methods. INL's EMRALD code performed dynamic PRA calculations of the identified sequences, integrating risk analysis with on-line deterministic safety analysis results (seismically-induced internal flooding calculations by NEUTRINO code and system analysis by RELAP5-3D code). RELAP5-3D Best-Estimate calculations coupled with RAVEN statistical code were used for deriving limit surfaces as function of relevant transient parameters. This data was used by EMRALD for determining the final core status (core damage/core safe). Finally, the coupling of seismic, flooding, system thermal-hydraulic, uncertainty quantification and probabilistic risk analysis for an advanced risk-informed external hazards analysis was demonstrated. Results showed the sensible decrease in the level of conservatism achieved using the developed risk-informed methodology, which allowed also improving the evaluation of the failure modes and failure probabilities of different components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

AUTOMATIC LIMIT SURFACE SEARCH FOR PWR TRANSIENTS BY RELAP5-3D/RAVEN CODES

Deterministic safety analysis of nuclear power plants can be improved by applying modern algorithms and codes exploiting machine learning and parallel computing. In this paper we show the application of some of these advanced methodologies for the determination of limit surfaces. A limit surface is an n-dimensional surface describing the plant status, e.g. identifying, as function of selected plant parameters, the boundaries between failed and safe conditions for the core fuel. Using the RELAP5-3D system thermal-hydraulic code coupled to the RAVEN statistical-analysis code, we show how to apply machine-learning algorithms (near-neighbor, support-vector machine, etc.) for determining relevant limit surfaces for PWR station-blackout and loss of offsite power. As a first step, a set of training simulations is run to sample the nuclear power plant responses. Those calculations are used to train a Reduced-Order-Model (ROM), which RAVEN then uses for determining an approximated limit surface. Verification of the guessed limit surface is then automatically performed and calculation is terminated depending by the chosen convergence criteria. The combination of High Performance Computing (HPC) and ROMs allow a remarkable reduction of the time required for the surface determination. RAVEN performs also the automatic data collection of the output of the numerous runs, thus easing the workload of the safety analyst. Finally, using RAVEN “Ensamble Forward” sampling technique, we show how to inform the limit surface calculations with some of the code and input uncertainties.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Energy Research at Idaho National Laboratory

Energy-related research at Idaho National Laboratory (INL) takes many forms. In this presentation, Liam Boire will step through a specific, on-going research project at INL which seeks to understand transmission grid resilience, and the components which can be improved to increase the resilience of our electrical infrastructure. While utilities and researchers similarly strive to improve power grid reliability, the frequency of large power outages are increasing. Thus, the importance for developing efficient methodologies that enable utilities and researchers alike to identify the most effective areas for improvement within the electric grid is more critical than ever. The method developed during this project recognized that a variety of factors were important to consider when identifying areas for improvement in the electric grid. Physics-based simulators are able to calculate voltage stability, load draws and generation capacities, and enumerate the combinations of component failures which result in system-wide failures (blackout scenarios). Probabilistic Risk Assessment (PRA) is able to assign probabilities to these failure scenarios based on typical component data (or more specific data if a specific system is being analyzed), and thus prioritize the list of failure scenarios based on likelihood and the components based on importance. Finally, adaptive capacity models can simulate the candidate grid and evaluate real and reactive power generation versus demand over time, calculating the total adaptive capacity available in various regions of the grid. The less adaptive capacity a region has, the less “buffer” that region has from failure. Not only are the individual results of these analyses valuable in understanding grid resilience, but the results of these analyses can be cross-correlated for deeper insight. When a grid is analyzed with this approach, failure scenarios can be enumerated and ranked by likelihood, components can be ranked by likelihood to drive system-wide failure, regions can be evaluated for adaptive capacity, and sensitive regions can be identified by recognizing the locations of important components within sensitive regions. Following this detailed discussion, a brief overview of some of the other energy research and areas of interest for the lab will be presented.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2020

Emergency Diesel Generator Backup Power Systems for Military Bases

Backup power systems on military installations must provide reliable power during a grid outage. The risks of blackouts and loss of electric power are not new. Outages of just a few hours are well known, but longer duration outages are becoming more frequent. The U.S. Army, Navy, and Air Force now require backup power from one to two weeks. For multiday outages, the reliability of emergency diesel generators will have a significant impact on the installation’s backup power system’s ability to provide power for critical missions.

backup power↗
DOE OSTI2020

Spatiotemporal Pattern Recognition in the PMU Signals in the WECC system

Phasor measurement unit (PMU) data has been used by multiple power system applications, including state estimation, post event analysis, oscillation detection, model validation, and many others. Still, due to its big data nature and availability to general research institutions, comprehensive understanding of the spatiotemporal patterns and underlying mechanisms are incomplete. This study applies a set of signal processing and machine learning approaches aiming at deciphering the characteristic behaviors of multiple phasor measurement units (PMUs) attributes (e.g., voltage, frequency, rate of change of frequency, phase angle), including their auto-correlation, cross-dependence, similarities and discrepancies across units and temporal scales, and distributions of anomalies and their linkages to potential external factors such as weather events. Data analytics are applied to PMUs from the U.S. Western Electricity Coordinating Council (WECC) system. The PMU measurements, recorded events, outages, and weather extremes are all from real world datasets. The findings from the study and mechanistic understanding of the PMU dynamics help provide guidance on system control or preventing blackouts. The derived metrics can be directly used for adjusting or filtering simulated PMU data used for advanced algorithm development.

Hou, Zhangshuan↗
DOE OSTI2020

Formulations and Valid Inequalities for Optimal Black Start Allocation in Power Systems

The restoration of a power system after a blackout starts around units with enhanced technical capabilities, referred to as black start units (BSUs). We examine the planning problem of optimally allocating these units on the grid subject to a budget constraint. We present a mixed integer programming model based on current literature in power systems. Binary variables are associated with the allocation of BSUs and with the energization state of buses, branches, and generators of the power system over a time horizon. We extract a substructure of the feasible region which imposes the requirement that each island that appears during the restoration process must have at least one operational generator. We discuss three equivalent reformulations for this requirement. We introduce a family of exponentially many, polynomially separable, valid inequalities to strengthen the formulation. Under simplifying assumptions, we show that the convex hull of the feasible region is a full-dimensional polyhedron, and prove that some of the constraints we introduced are facet-defining. We perform experiments to examine the difference in strength between the formulations as well as the computational times to solve the problem to near optimality for synthetic instances of the IEEE-39, IEEE-118, Illinois-200, WECC-225, IEEE-300, South Carolina-500, and Texas-2000 power systems. We illustrate a use case of the model. We conclude by suggesting extensions of the current work for future research.

Black Start Allocation↗
DOE OSTI2021

Systems and methods for controlling electrical grid resources

The present invention relates to PMU-based control systems for dampening inter-area oscillations in large-scale interconnected power systems or grids to protect against a catastrophic blackout. The control systems receive phasor measurements from two or more locations on an AC transmission line and generates a power control command to a power resource on the grid.

24 POWER TRANSMISSION AND DISTRIBUTION↗
DOE OSTI2021

A Commercial Building Plug Load Management System that Uses Internet of Things Technology to Automatically Identify Plugged-In Devices and Their Locations

Plug and process loads (PPLs) account for a large portion of U.S. commercial building energy use. There is a huge potential to reduce whole building consumption by targeting PPLs for energy savings measures or implementing some form of plug load management (PLM). Despite this potential, there has yet to be a widely adopted commercial PLM technology. This paper describes the Automatic Type and Location Identification System (ATLIS), a PLM system framework with automatic and dynamic load detection (ADLD). ADLD gives PLM systems the ability to automatically identify devices as they are plugged into the outlets of a building. The ATLIS framework takes advantage of smart, connected devices to identify device locations in a building, meter and control their power, and communicate this information to a central database. ATLIS includes five primary capabilities: location identification, communication, control, energy metering, and data storage. A laboratory proof of concept (PoC) demonstrated all but the energy metering capability, and these capabilities were validated using a series of system tests. The PoC was able to identify when a device was plugged into an outlet and the location of the device in the building. When a device was moved, the PoC's dashboard and database were automatically updated with the new location. The PoC implemented controls to devices from the system dashboard so that devices maintained correct schedules regardless of where they were plugged in within the building. ATLIS's primary technology application is improved PLM, but other applications include asset management, energy audits, and interoperability for grid-interactive efficient buildings. An ATLIS-based system could also be used to direct power to critical devices, such as ventilators, during a brownout or blackout. Such a framework is an opportunity to make PLM more widespread and reduce the amount of energy consumed by PPLs in current and future commercial buildings.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2022

Assessment of Thermal Stratification in Versatile Test Reactor Transients

The Versatile Test Reactor (VTR) is a fast-spectrum test reactor currently being developed in the United States under the direction of the US Department of Energy. The conceptual design of the 300 MWth pool-type sodium-cooled fast reactor (SFR) has been led by the US National Laboratories in collaboration with General Electric-Hitachi and Bechtel National Inc. Safety performance analysis for the VTR conceptual design is being performed with the systems thermal-hydraulics (SYSTH) module of the SAS4A/SASSYS-1 liquid-metal reactor safety analysis code system. Since the current model of the VTR employs a simple perfect mixing model for large plena like the hot pool, it is not able to predict temperature variations that may develop during the transient. Prior work simulating the response of SFRs to postulated events like the Protected Station Blackout (PSBO) has shown that the phenomenon of thermal stratification, where stable thermal layers accumulate in the hot pool, may delay the transition to natural circulation and thus impact the predicted transient progression. Thus, an effort has begun to model this transient by integrating a Computational Fluid Dynamics (CFD) model of the hot pool into the SAS4A/SASSYS-1 model of the Primary Heat Transport System during the simulation of the PSBO event. A three-dimensional Volume-Of-Fluid CFD model of the VTR hot pool has been developed for the co-simulation of SAS4A/SASSYS-1 with CFD. In this work, the standalone SAS4A/SASSYS-1 calculation and the standalone CFD calculation based on the SAS4A/SASSYS-1 calculation result were produced on the Idaho National Laboratory High Performance Computing cluster, SAWTOOTH. At this time, only the standalone CFD and SAS4A/SASYS-1 simulation results are provided.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

A Systematic Study to Determine 5G Baseline Performance for Scientific Computing

The fifth-generation (5G) cellular networks envisions achieving higher data rates, improved connectivity, reduced latency, and better quality of service (QoS) than the fourthgeneration (4G) cellular networks. Such improved performance can be utilized to address the challenges in applications such as electricity generation in power systems. The traditional power grids responsible for electricity generation suffer from drawbacks such as life-threatening blackout crises, and energy storage proliferation as they are not robust to extreme climatic conditions. A recent study proposed the idea of extending the capabilities of advanced wireless technologies such as the current 5G to develop a robust, energy-efficient, and secure smart grids. However there are two main challenges associated with the integration of power systems and wireless technologies. First, it is imperative to understand the architecture and the enabling technologies of 5G to ensure that the performance requirements of the smart grids are met. Second, an end-to-end testbed is required to determine if the performance requirements are met by estimating the 5G characteristics such as latency, and throughput. Our proposed alleviates the aforementioned concerns in the following manner. To begin with, a systematic study of the 5G architecture including both the StandAlone (SA) and Non-Standalone (NSA) operations is presented. Furthermore, a detailed survey of the possible 5G enabling technologies is elicited. In addition to these, an end-toend testbed that can estimate the 5G characteristics is explained in detail with appropriate preliminary results.

5G, 5G Communication↗
DOE OSTI2021

METHODOLOGY AND APPLICATION OF PHYSICAL SECURITY EFFECTIVENESS BASED ON DYNAMIC FORCE-ON-FORCE MODELING

This paper describes ongoing work within the Light Water Reactor Sustainability (LWRS) Program at Idaho National Laboratory (INL) to optimize security and cost of nuclear power plants (NPPs). It reviews the conservatisms in conventional physical security posture and regulations. It introduces the dynamic risk assessment tool developed at INL, Event Modeling Risk Assessment using Linked Diagrams (EMRALD). The dynamic assessment methodology leverages EMRALD to process results of force-on-force (FOF) simulations and crediting safety mitigation actions from probabilistic risk assessment (PRA) models as well as diverse and flexible coping strategies (FLEX) mitigation strategies. Timing information from these simulations are compared against the available time to perform mitigations obtained from Reactor Excursion and Leak Analysis Program (RELAP5) simulations. To illustrate the methodology, a station blackout (SBO) attack scenario was modeled in commercially available FOF simulation tools. The simulation results provide valuable insights into possible attack outcomes and as the probabilistic risk of a core damage event given these outcomes. Safety mitigation procedures were modeled in EMRALD, and were dependent on the attack outcomes by considering human operator uncertainties. RELAP5 simulations incorporating human and hardware uncertainties were performed to estimate the distribution of time-to-core damage. The results demonstrate that, even in the extreme case of a successful adversarial attack, plant mitigation strategies provide significantly high-likelihood of preventing radiological release. The proposed modeling and simulation framework of integrating FLEX equipment with FOF models enables the NPPs to credit FLEX portable equipment in the plant security posture, resulting in an efficient and optimized physical security.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

AP1000{sup R} nuclear power plant safety overview for spent fuel cooling

The AP1000{sup R} plant is an 1100-MWe class pressurized water reactor with passive safety features and extensive plant simplifications that enhance construction, operation, maintenance, safety and costs. The AP1000 design uses passive features to mitigate design basis accidents. The passive safety systems are designed to function without safety-grade support systems such as AC power, component cooling water, service water or HVAC. Furthermore, these passive features 'fail safe' during a non-LOCA event such that DC power and instrumentation are not required. The AP1000 also has simple, active, defense-in-depth systems to support normal plant operations. These active systems provide the first level of defense against more probable events and they provide investment protection, reduce the demands on the passive features and support the probabilistic risk assessment. The AP1000 passive safety approach allows the plant to achieve and maintain safe shutdown in case of an accident for 72 hours without operator action, meeting the expectations provided in the U.S. Utility Requirement Document and the European Utility Requirements for passive plants. Limited operator actions are required to maintain safe conditions in the spent fuel pool via passive means. In line with the AP1000 approach to safety described above, the AP1000 plant design features multiple, diverse lines of defense to ensure spent fuel cooling can be maintained for design-basis events and beyond design-basis accidents. During normal and abnormal conditions, defense-in-depth and other systems provide highly reliable spent fuel pool cooling. They rely on off-site AC power or the on-site standby diesel generators. For unlikely design basis events with an extended loss of AC power (i.e., station blackout) or loss of heat sink or both, spent fuel cooling can still be provided indefinitely: - Passive systems, requiring minimal or no operator actions, are sufficient for at least 72 hours under all possible pool heat load conditions. - After 3 days, several different means are provided to continue spent fuel cooling using installed plant equipment as well as off-site equipment with built-in connections. Even for beyond design basis accidents with postulated pool damage and multiple failures in the passive safety-related systems and in the defense-in-depth active systems, the AP1000 multiple spent fuel pool spray and fill systems provide additional lines of defense to prevent spent fuel damage. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Evaluation of earthquake and tsunami on JSFR

Evaluation of earthquake and tsunami on JSFR has been analyzed. For seismic design, safety components are confirmed to maintain their functions even against recent strong earthquakes. As for Tsunami, some parts of reactor building might be submerged including component cooling water system whose final heat sink is sea water. However, in the JSFR design, safety grade components are independent from component cooling water system (CCWS). The JSFR emergency power supply adopts a gas turbine system with air cooling, since JSFR does not basically require quick start-up of the emergency power supply thanks to the natural convection DHRS. Even in case of long station blackout, the DHRS could be activated by emergency batteries or manually and be operated continuously by natural convection. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗