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The present invention relates to surface mount structures including a capacitive element or a resistive element, where the element has a property that is responsive to an environmental condition. In particular examples, the structure can be optionally coupled to a printed circuit board. Other apparatuses, surface mountable structures, and methods of use are described herein.
Supply chain security to detect, deter, and prevent the counterfeiting of networked and stand-alone integrated circuits (ICs) is critical to cyber security. Sandia National Laboratory researchers have developed IC ID to leverage Physically Unclonable Functions (PUFs) and strong cryptographic authentication to create a unique fingerprint for each integrated circuit. IC ID assures the authenticity of ICs to prevent tampering or malicious substitution.
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In analog instrumentation and control (I and C) systems for nuclear power plants (NPPs), most connections are comprised of hardwired point-to-point types. Since they are based on analog technologies, they are fragile to noise and require analog-to-digital conversion time. Therefore, many NPPs have adopted communication network for the exchange of safety-critical data between programmable logic controllers (PLCs) in digital I and C systems to effectively accommodate a huge number of field controllers. However, the primary issue of digital network communication in a safety-critical system involves potential hazards, such as partial or total loss of communication between PLCs, which may result in a failure of safety critical signal generation when it is needed. Therefore, the probability that a safety-critical system in digitalized NPP becomes unsafe due to a network failure must be evaluated to quantify the risk of digital I and C system. In Korea, the Engineered Safety Feature-Component Control System (ESF-CCS) was developed as one of the digital safety-critical system in Advanced Power Reactor- 1400 (APR-1400). The function of a digital engineered safety feature (ESF) actuation system and field actuator controllers of conventional NPPs are performed by the ESF-CCS. In this system, safety-critical network, High Reliable- Safety Data Network (HR-SDN), was employed for the safety-critical data transmission from group controllers (GCs) to loop controllers (LCs). In this study, a framework for assessing the reliability of a HR-SDN network communication between GCs and LCs in the ESF-CCS in addition to a fault-tree model for ESF components considering network failure between GCs and LCs were developed to assess the plant risk effect of a newly developed digitalized ESF-CCS. Based on the case studies and cut set analysis, the risk effect of network failure on plant risk and the important risk contributors of network failure were quantitatively addressed. A framework for identifying the potential hazardous states of network communication in the ESFCCS was proposed. In addition, a fault-tree model for network communication failure was developed to estimate the risk effects of network failure between the GCs and LCs on ESF-CCS signal failure. The developed fault-tree model was then applied to several case studies. The fault-tree model of ESF-CCS signal failure for ESF components required for various mitigation actions regarding NPP DBAs was designed by considering the identified hazardous states of network failure that would result in a failure to provide input signals to the corresponding LC. As a sensitivity study, three case studies were performed based on the periodic test methods for hardware components in HR-SDN module and their test intervals in this study. The quantitative results for three case studies demonstrated that the probability of overall network communication failure, which was calculated as the sum of the failure probability associated with each failure cause, contributed up to 1.58% on the Small-LOCA CD frequency. The results also revealed the important basic events related to network failure in the case studies which included CCF of hardware and software in HR-SDN module. This study is expected to provide insight into the development of the fault-tree model for the network failure in digital I and C system and into the quantification of the risk effect of network failure for safety-critical data transmission in NPPs. As a further study, the risk effects of network failure in the ESF-CCS on the plant risk can be estimated in a more detailed manner by analyzing the failure modes and causes in a sub-component level and the fault detection coverage of the fault tolerance techniques for network failure implemented in ESF-CCS after the detailed configuration of ESF-CCS is decided. In terms of the failure of manual ESF actuation, the relationship among the human action failures must be investigated to consider multiple human error condition since ESF-CCS provides multiple manual actuation to assure the diversity of the system. Since the conventional HEP method cannot accommodate the multiple conditions in a fault tree, condition-based human reliability assessment (CBHRA) method or the dynamic human reliability modeling approach can be applied to include a complex relationship among the automated safety signal generation and the human operator's manual actuation, avoiding the optimistic estimation of HEP of the human action failure.
During planned, periodic outages, a selection of fuel channels within the UK fleet of Advanced Gas-cooled Reactor (AGR) cores are inspected using specialist tools which record video footage and other sensory data for each channel which undergoes inspection. Current visualization techniques comprise of manually produced montages by inspection engineers of points of interest (i.e. structural defects) and 2-D panoramic images of the fuel channels automatically produced using bespoke image stitching software. Both techniques however provide limited structural information due to the loss of depth data as a result of the image formation process. By recovering the depth information from the footage, a 3D model could be constructed and subsequently, allow for more accurate profiling of specific defects observed during inspection in addition to obtaining the fuel channels structure using existing footage. This work explores the preliminary application of a 3-D visualization technique known as Structure-from-Motion (SfM) which aims to obtain 3-D information by exploiting image correspondences across multiple viewpoints of the same scene in the RVI footage. This paper investigates the difficulties of applying state-of-the-art SfM to RVI footage and we present new techniques to improve feature correspondence searching in repetitive, nondescript environments. (authors)
The final draft of a revision to IEEE Std. 1012-2012, 'IEEE Standard for System, Software and Hardware Verification and Validation', is expected to be approved for publication in June 2017. A draft of NEI 96-07, Appendix D, 'Supplemental Guidance for Application of 10 CFR 50.59 to Digital Modifications', has been submitted to the NRC staff for review and endorsement. What, if any, relationship exists between these two documents? How can the nuclear industry utilize the first one to help ensure success when applying the second one? NRC Regulatory Guides only endorse digital system and software related industry standards for application to safety-related systems and components, but the vast majority of digital modifications are made to non-safety related systems and components, so are these industry standards of little or no use when performing digital upgrades to non-safety related equipment? This paper answers these and other related questions and shows how application of a graded approach to verification and validation, utilizing the IEEE Std. 1012 concept of 'integrity level', can increase the likelihood of a 10 CFR 50.59 evaluation outcome that does not require prior NRC review and approval for non-safety related digital modifications. (authors)
Lack of appropriate human performance data for the development of human reliability analysis (HRA) methods has been identified as a key issue contributing to the method-to-method variability among the HRA methods. The United States Nuclear Regulatory Commission (NRC) worked with domain experts to develop the Scenario Authoring, Characterization, and Debriefing Application (SACADA) system to collect nuclear power plants' licensed operator simulator training and examination data to inform HRA. In the past two decades, significant improvements in understanding and modelling human performance in HRA have affected the perspectives of the types of data to be collected for HRA. A significant change is that human performance is strongly dependent on the context of the tasks to be performed. To incorporate this understanding, the SACADA data taxonomy collects the context and actual performance information of the training objective elements (TOEs) in nuclear plant operator simulator training. This provides a framework consistent with HRA to use context to estimate human error probabilities. It is recognized that measuring crew performance is equally important as improving operator training. SACADA is designed to support the nuclear trainers in capturing this important information that is used to identify training weaknesses and drive continuous improvement. This paper discusses the SACADA taxonomy, the SACADA system's functions, and the supporting software design to enable the SACADA system to be practically implemented in the nuclear power plants' operator simulator training program to collect data for tracking operator performance and HRA analysis. (authors)
Unscheduled shutdown of nuclear power facilities for recalibration and replacement of faulty sensors can be expensive and disruptive to grid management. In this work, we present virtual (software) sensors that can replace a faulty physical sensor for a short duration thus allowing recalibration to be safely deferred to a later time. The virtual sensor model uses a Gaussian process model to process input data from redundant and other nearby sensors. Predicted data includes uncertainty bounds including spatial association uncertainty and measurement noise and error. Using data from an instrumented cooling water flow loop test bed, the virtual sensor model has predicted correct sensor measurements and the associated error corresponding to a faulty sensor. (authors)
Very few nuclear power plants are using computerized procedures to guide the control room operation. A CPS implementation consists of two processes: digitalization and computerization. The former deals with the transformation of paper-based procedure into a digital data; the latter is referred to as the addition to that data of extra information that is meaningful for decision-making. This information may include real-time data (plant parameter values and binary process values to indicate equipment states), provide access to trends and historical data, etc. In order to be efficient and to optimize the digitalization and computerization processes, these set of aids should be implemented at the same time as operating procedures are digitalized There are references of plants having successfully implemented Computerized Procedures Systems (CPS) for emergency procedures, but when questioned about normal or abnormal operating procedures, operators and plant management seem reluctant to take that step. Given that a CPS is meant, by design concept, to increase operation crew efficiency by facilitating the procedures execution and reducing human error, why are we reluctant to enjoy CPS benefits in the most common plant status (i.e. normal operation)? Although there are many answers to this question, one is shared by most of the users: CPSs are not designed for editing and maintaining procedures easily and reliably. Normal operating procedures, in opposition to the emergency procedures, are often reviewed. Nowadays, these edition activities are performed easily with a word processor software and using paper, but what is required when dealing with a CPS? (authors)
A new conceptual frame of how to design and validate a digital HIS (human interface system) on an innovative numerical simulation basis is proposed for the support of plant operators' supervisory control of various types of automated complex NPPs (nuclear power plants). The proposed conceptual framework utilizes the object-oriented software for plant DiD (defense-in depth) risk monitor with the combination of nuclear reactor accident simulation by an advanced nuclear safety analysis code RELAP5/MOD4. The conceptual frame proposed in this paper will be applied for an example practice for the SBLOCA (small break loss of coolant accident) case of passive safety PWR (pressurized water reactor) AP1000. (authors)
Refueling outages remain one of the largest opportunities for improving capacity factor and reducing costs available to commercial nuclear power plants. Although the nuclear industry has made steady improvement in outage optimization, each day of a refueling outage still represents an opportunity to save millions of dollars and each day an outage extends past its planned end date represents millions of dollars that may have been spent unnecessarily. Reducing planned outage duration or preventing outage extensions requires careful management of the outage schedule as well as constant oversight and monitoring of work completion during the outage execution. During a typical outage, there are typically more than 10,000 activities on the schedule that, if not managed efficiently, may cause expensive outage delays. Management of outages currently relies largely on paper-based resources and general-purpose office software. A typical method currently used to monitor work performance is a burn-down curve, where total remaining activities are plotted against the baseline schedule to track bulk work completion progress. While methods like this are useful, there is still considerable uncertainty during a typical outage whether bulk work progress is adequate and therefore a lot of management time is spent analyzing the situation on a daily basis. This paper is a case study of how new technology, in combination with established human-factors and user-centered design knowledge, can be used to support Outage Control Center (OCC) personnel decision making. In particular, we describe recent advances made in developing a framework for the design of visual outage information presentation, as well as an overview of the human factors principles that informed the development of the visualizations. To test the utility of advanced visual outage information presentation, an outage management dashboard software application was created as part of the Department of Energy's Advanced Outage Control Center project (AOCC). This dashboard is intended to present all the critical information an outage manager would need to understand the current status of a refueling outage. The dashboard presents the critical path, bulk work performance, key performance indicators, outage milestones and metrics relating current performance to historical performance. Additionally, the dashboard includes data analysis tools to allow outage managers to drill down into the underlying data to understand the drivers of the indicators. (authors)
Digitalization of the instrumentation and control systems in nuclear power plant entails some new features which do not exist in analog systems. When the new features are applied in safety critical systems, the risk stemming from them should be quantified properly to assure the reliability of the entire power plant. Among the new features, software is considered as the most important factor because it can cause common cause failure to many redundant systems. Therefore, in this work, a test based reliability assessment method for a safety-critical software is developed. An output of a software is determined by not only inputs but also the internal state of the software at the time. In the case of safety critical software developed for usage in nuclear power plant, it is possible to directly investigate probable internal states thanks to detailed design specification and programing features, and the probable inputs also can be identified based on the physical linearity of each process parameter and hardware characteristics related to data acquisition. In this work, a development method for practical exhaustive test case consisting of investigation of internal state and probable input sets to a specific internal state are suggested. In addition to this, software logic simulator for execution of software logic test and the reliability quantification method based on the test result are developed. The feasibility of the suggested method is demonstrated via a case study. (authors)
Redundant measurement, trip, and actuation channels are required for reactor trips and engineered safety systems in nuclear power plants. There are three or four independent channels in existing plants, depending on the system. To prevent failures, consistent signals from any two channels together can trip the reactor or actuate safety systems. Measurements associated with safety trips and actuations include uncertainties related to the instruments making the measurements and to process conditions. Margins that include measurement uncertainties are set for each trip and actuation channel as though it were the only one taking action, thus preserving independence. There are sensors, however, that provide partially redundant information that could be used to decrease the uncertainty associated with measurements of plant conditions, while maintaining the three-channel redundancy important to plant safety. Reducing these uncertainties would both improve plant safety by providing more accurate knowledge of plant conditions and improve the operating margins that must take such uncertainties into account. This paper demonstrates how information that is already available to plant instrumentation and control systems and is already redundant can be used through software routines to reduce the error that must be taken into account in reactor trip and safety system actuation setpoints. Issues that would have to be addressed related to standards, licensing, and implementation are also discussed. (authors)
Nuclear power plants are increasingly adding digital components for plant operation, safety, and security. These digital components fill a gap with legacy equipment where replacement components no longer exist. They also benefit operation of the plant by increasing efficiency in power generation, monitoring of equipment and plant parameters, as well as aiding operator control. However, the addition of digital components and systems also adds cyber risks with previously unanalyzed failure modes and attack vectors are introduced with these new systems. These risks are difficult to identify, analyze, and mitigate due to the increasingly complex nature of the digital components and the integration of these components with additional plant processes and communication networks. The research presented in this paper develops a new method that addresses the cyber risk to inform appropriate levels of protection. EPRI and Sandia are working under a Cooperative Research and Development Agreement to develop an effective method of evaluating the cyber risk in production nuclear power facilities. The Cyber Hazards Analysis Risk Methodology (CHARM) focuses on ensuring adequate controls are in place for appropriate cyber protection of the plant from radiological release or generation risk. Existing plant hazards analyses (e.g., PRA, FTA) do not account for software deficiencies or adversarial intent. This method leverages existing plant analyses and MIT's Systems Theoretic Process Analysis (STPA) to create cyber informed fault trees. These new fault trees will provide the basis for comprehensive cyber risk analysis and help ensure potential gaps in cyber security controls are identified and corrected. (authors)
The control systems of nuclear power plants contain many components which are becoming obsolete at an increasing rate. Nowadays maintenance for analog-based protection systems may be difficult as analog parts become obsolete or hard to obtain. Various studies have been conducted to address this control system hardware obsolescence issue. Those control systems of nuclear power plants have been replaced with modern digital control systems such as the programmable logic controller (PLC) and the distributed control system (DCS). The central processing unit (CPU)- based platforms such as the PLC and the DCS require an operating system (OS) and application software, and thus they might result in a common-cause failure when a problem occurs in the operating system or application software. The field programmable gate array (FPGA) is highlighted as an alternative to the conventional protection or control systems. This paper presents the development of an FPGA-based logic controller (FLC) and a diverse protection system (DPS) with the FLC platform. Protection functions for the DPS were implemented in FPGAs without any CPU or OS. (authors)
In February 2016, Nuclear Energy Institute (NEI) launched its 'Delivering the Nuclear Promise (DNP)' strategic plan with the objective of reducing electric generating costs by 30% across the nuclear power industry by 2018. Improving efficiency has been addressed as one of the strategic focus areas in the plan. In nuclear power plants (NPPs) there is room for improvement in main components efficiency and TecSOLCEP is a software tool that has proved its value for that purpose. TecSOLCEP is a software tool for monitoring, diagnosing and quantifying performance deviations in Power Plants. It faithfully follows the methodology of ASME standards. Over 15 years of success stories in more than 50 electric power plants of different technologies worldwide, justify the benefits for the plants. The implicit concept in TecSOLCEP is 'Benefit based Maintenance' (BbM) as complementary to usual maintenance strategies like time-based, condition-based or predictive. These maintenance strategies are oriented to obtain the Safety and Reliability goals shared by all the electricity generation technologies. BbM is oriented to get the maximum economic benefit by Efficiency care, which historically has not been a priority, especially in nuclear plants. BbM applied with the help of performance monitoring tools like TecSOLCEP, is called to play an important role to improve efficiency which, in plants mainly operating at full power, is equivalent to increase the maximum power. Nuclear power plants that intend to extend their lives may also take advantage of identifying and quantifying inefficiencies in equipment to better address necessary investments. (authors)
Obsolescence, aging, reliability, and performance issues are driving nuclear facilities to replace conventional analog Instrumentation and Control (I and C) systems with digital technologies. In addition, the designs of the next generation of nuclear reactors, including Small Modular Reactors (SMRs), will incorporate digital I and C for most, if not all, of their safety and non-safety related functions. As part of an on-going research and development (R and D) effort under the auspices of the U.S. Department of Energy (DOE), Analysis and Measurement Services Corporation (AMS) is developing a platform to provide automated testing of digital I and C systems and create a standard method of evaluation for reliability assessments. This hands-on R and D effort has produced a Software Reliability Tester (SRT), which is a set of software and hardware tools designed to automate the testing of digital I and C systems to measure and quantify how well the systems perform under normal operating conditions and in the presence of faults. Furthermore, the SRT provides the foundation for a practical tool to automate verification and validation (V and V) activities and reduce the amount of testing time of digital I and C systems. When combined with its capabilities of integrating both reliability and fault tolerance quantification, the SRT can be used to ensure that digital I and C implementations are both safe and cost-effective for the nuclear industry. This paper describes the application of the SRT to perform V and V of a new Digital Rod Position Indication (DRPI) diagnostic system. The DRPI diagnostic system is a digital system that monitors the rod address and rod position data bus voltages of typical DRPI systems in nuclear power plants to detect and diagnose faults. For this application, the SRT was configured to exercise the inputs of the DRPI diagnostic system, and test cases were generated for common system faults such as address and data bit errors, bus voltage, and timing problems. These test cases were applied by the SRT hardware as voltage inputs to the DRPI diagnostic system and the outputs were compared to expected values generated by a model of the DRPI diagnostic system. Included in the paper is a description of the overall design of the SRT including the hardware and software architectures. Methodologies for automating test cases are also described. V and V using the SRT demonstrates the benefits of automated testing and qualification to provide a quantitative assessment of reliability and cost effective implementation of digital I and C in existing and next generation nuclear power plants. (authors)
This paper studies the propagation and effects of faults of critical components that pertain to the secondary loop of a nuclear power plant found in Nuclear Hybrid Energy Systems (NHES). This information is used to design an on-line monitoring (OLM) system which is capable of detecting and forecasting faults that are likely to occur during NHES operation. In this research, the causes, features, and effects of possible faults are investigated by simulating the propagation of faults in the secondary loop. The simulation is accomplished by using the Integrated System Failure Analysis (ISFA). ISFA is used for analyzing hardware and software faults during the conceptual design phase. In this paper, the models of system components required by ISFA are initially constructed. Then, the fault propagation analysis is implemented, which is conducted under the bounds set by acceptance criteria derived from the design of an OLM system. The result of the fault simulation is utilized to build a database for fault detection and diagnosis, provide preventive measures, and propose an optimization plan for the OLM system. (authors)