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At least 343 records · Page 19
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DOE OSTI2017

Computerized procedure systems: procedure edition and maintenance challenges - 232

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

The problem analysis of data processing performance in nuclear power plant computerized information and control system - 288

This paper is based on the data processing performance problems of operating command delay and system display not updating real-time for nuclear power plant computerized information and control system. The plant computerized information and control system is responsible for monitoring and controlling the plant operation, and it is a key part in the plant DCS system. Its data processing performance will affect the operation stability of nuclear power plant. The architecture, hardware configuration and data communication for plant computerized information and control system among the different levels in DCS are described. The main reasons affecting data processing performance in different signal channels are analyzed according to the plant load test results. The CPU load curves for real time server and central processing server are displayed, although the detailed cause for operating command delay and system display not updating realtime are not found, central processing server load is heavier than real time server which remind us to analyze the data processing of software resource scheduling deeply. So the resource distribution problem of software execution task is also analyzed according to CPU load test results. This paper provides the optimization analysis for data processing performance, and it is available and effective according to the latest test results. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Human factors and modeling methods in the development of control room modernization concepts - 315

Control room modernization is one of the most challenging and complex upgrade projects that a nuclear power plant can undertake. It can even have a bigger impact on operations than, for example, turbine replacement. The challenges of migrating an analog control system to a distributed control system are already well known and a number of nuclear utilities have embarked upon various levels of effort to upgrade some of the systems in the control room. When planning for control room upgrades, plants have to deal with a multitude of engineering and operational impacts. This will inevitably include several human factors considerations, including physical ergonomics of workstations, viewing angles, lighting, seating, new communication requirements, and new concepts of operation. In helping nuclear power utilities to deal with these challenges, the Idaho National Laboratory (INL) has developed human factors design and evaluation methods to be used in the development of an end-state control room concept and to manage the various phases of the upgrade life cycle. This included interactive sessions with operators in INL's Human System Simulation Laboratory, and three-dimensional modeling to visualize control board changes. To make the various upgrade phases as concrete and as visible as possible, the end-state concept includes a set of visual representations of the control room before and after various upgrade phases to provide the context and a framework within which to consider the various options in the upgrade. This includes the various control systems, human-system interfaces to be replaced, and possible changes to operator workstations. This paper describes how this framework helps to ensure an integrated and cohesive outcome that is consistent with human factors engineering principles and also provide substantial improvement in operator performance. The paper further describes the application of this integrated approach in the strategic modernization program at a nuclear power plant where legacy systems are upgraded to advanced digital technologies through a systematic process that links human factors principles to the systems engineering process. This approach results in an integrated control room architecture beyond what is possible for individual subsystem upgrades alone. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

A novel technique applying spectral estimation to Johnson noise thermometry - 317

Johnson noise thermometry is one of many important measurement techniques used to monitor the safety levels and stability in a nuclear reactor. However, this measurement is very dependent on the minimal electromagnetic environment. Properly removing unwanted electromagnetic interference (EMI) is critical for accurate drift free temperature measurements. The two techniques developed by Oak Ridge National Laboratory to remove transient and periodic EMI are briefly discussed in this paper. Spectral estimation is a key component in the signal processing algorithm used for EMI removal and temperature calculation. The cross-power spectral density is a key component in the Johnson noise temperature computation. Applying either technique requires the simple addition of electronics and signal processing to existing resistive thermometers. With minimal installation changes, the system discussed here can be installed on existing nuclear power plants. The Johnson noise system developed is tested at three locations: Oak Ridge National Laboratory, Sandia National Laboratory, and Tennessee Valley Authority (TVA) Kingston Steam Plant. Each of these locations enabled improvement on the EMI removal algorithm. The conclusions made from the results at each of these locations is discussed, as well as, possible future work. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Fault propagation and effects analysis for designing an online monitoring system for the secondary loop of a nuclear power plant part of a hybrid energy system - 393

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Embedded Sensors and Controls to Improve Component Performance and Reliability - Final Report

Oak Ridge National Laboratory has researched embedded instrumentation and controls (I&C) for extreme environments under the Department of Energy, Office of Nuclear Energy's Nuclear Energy Enabling Technology program. Advanced sensors and controls have the potential to dramatically increase the performance and efficiency of nuclear reactor components, however, the extreme environments of exiting and advanced reactor designs limit the options for sensors, actuators, and control designs. To address two of the extreme environment challenges to embedded I&C, the focus of this research is to fill technology gaps needed for electromagnetic systems that can operate at temperatures greater than 700 C in corrosive environments. A magnetically suspended, canned-rotor pump was chosen for the embedded I&C development platform because it is a system with a practical need in industry and because the sensors, actuators, and embedded I&C technologies needed to successfully build this platform will be widely applicable to any high-temperature components that utilize electromagnetic interactions. This design also eliminates the need for pump seals and mechanical bearings which are the main sources of pump failures and maintenance costs. During the project a conceptual design for the high-temperature magnetically suspended pump was developed and analyzed for manufacturability and reliability. Specifications for the materials, control system, sensors, mechanical performance, high-temperature wiring, and electromagnetic properties were developed for the conceptual design. The canned rotor design of the high-temperature pump presents unique fluid force characteristics arising from the thin fluid channel between the rotor and stator. A model of the fluid forces was developed and the impact of these forces on the stability and performance of magnetic bearing controllers was studied. Then a room temperature bench-scale magnetic testbed was designed and fabricated for testing the performance of sensors and electromagnetic actuators, modern control design techniques, closed-loop system identification, state estimation, and model uncertainty. Finally, a loop-scale testbed with submerged magnetic bearings was designed and fabricated. This testbed will be used to analyze the stability and performance of control system designs in response to the nonlinear and cross-coupling fluid effects between the shaft axes of motion, rotordynamics, gyroscopic effects, and impeller disturbances.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Virtual sensors for robust on-line monitoring (OLM) and diagnostics - 197

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Enhancing workload assessments for validation activities associated with DBA and BDBA scenarios - 165

After the Fukushima Daiichi accident, nuclear regulators around the world have required that power reactor licensees develop more extensive emergency mitigating responses and severe accident management provisions beyond the defense-in-depth measures for design basis accidents (DBAs), previously in place. Workload assessments represent common validation techniques that are used to demonstrate that workers are able to perform tasks without unacceptable performance degradation. High workload is known to induced stress and fatigue and may severely diminish a worker's capacity to perceive, recognize, and respond appropriately during emergency or unanticipated events, which may result in undesirable consequences. In estimating workload during emergency and severe accident scenarios, power reactor licensees tend to rely on subjective measures of workload, such as NASA-TLX. Due to reported mismatches in the literature between subjective and physiologically-derived estimates of workload, it is prudent to see what more can be done to improve the current state of practice, in the context of emergency and severe accident conditions. To improve confidence in workload estimates, it is advocated that the nuclear industry integrate physiological-based measures into current practices by making use of 'on-body' or 'wearable' physiological sensors. In this paper, an overview of three different types of empirical workload approaches is provided. The advantages of wearable physiological sensors are considered in the context of extreme environments and occupations, with tangible examples including heat stress and pupillometry. Suggestions for a consensus forum on workload are provided and a research plan directed at improving the current practice of workload estimation is offered for consideration. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Utilizing the SACADA system to collect simulator training data to inform human reliability analyses - 169

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Design and implementation of HuREX analysis supporting interface - 171

Human error is recognized as one of the main causes to affect to the socio-technical system including nuclear power plant. Therefore, reducing human error through human reliability analysis (HRA) is a critical issue. Moreover, sufficient and reliable human performance data collection is a prerequisite for ensuring the safety of nuclear power plants. Korea Atomic Energy Research Institute developed the HuREX (Human Reliability data Extraction) framework to collect and analyze the responses of human operators from simulator training records of nuclear power plant. Thus, the development and computer software interfaces of the database are essential to efficiently support human reliability analysis. In this study, OPERA database was developed in accordance with the following process. First, the essential data required for the HRA analysis (e.g., plant data, procedures, system and equipment info, operating crews, results of simulator experiments, and so forth) was stored in the database. In addition, standardized procedures and related items were stored in the database to support HRA quantification. A data structure was developed to store crew responses conducted by operators to cope with an emergency situation during simulator experiments. In addition, various interfaces that escalate efficiency of the HuREX applications were developed. The HuREX analysis support interface was developed with HuREX Video Analyzer for observation of simulator collection data (e.g., audio-visual records, plant parameters, operator's action logs) and HuREX Data Analyzer for analysis of operator's conducted task. By using the OPERA database and HuREX analysis support interface developed through this research, maximizing the efficiency of human reliability analysis based on simulator training data. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Enhancing power plant safety through simulated cyber events - 341

There are gaps in understanding how a cyber-attack would manifest itself within power plants and what these events would look like within the control room from an operator's perspective. This is especially true for nuclear power plants where safety has much broader consequences than nonnuclear plants. The operating and emergency procedures that operators currently use are likely inadequate for targeted cyber-attacks. This research focuses on understanding how a cyber event would affect the operation of the plant, how an operator would perceive the event, and if the operator's actions would keep the plant in a safe condition. This research is part of Sandia's Laboratory Directed Research and Development program where a nuclear power plant cyber model of the control system digital architecture is coupled with a generic pressurized water reactor plant training simulator. Cyber event scenarios will be performed on the coupled system with plant operators. The scenarios simulate plant conditions that may exist during a cyber-attack, component failure, or insider sabotage, and provide an understanding of the displayed information and the actual plant conditions. These scenarios will determine if plant operators can 1) recognize that they are under cyber-attack and 2) take appropriate actions to keep the plant safe. This will also provide the opportunity to assess the operator cognitive workload during such events and identify where improvements might be made. Experiments with nuclear power plant operators will be carried out over FY 2018 and results of the research are expected by the end of FY 2018. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

An Approach for Assessing Consequences of Potential Supply Chain and Insider Contributed Cyber Attacks on Nuclear Power Plants

The Stuxnet attack at the Natanz facility is an example of a targeted and successful cyber attack on a nuclear facility. Snowden's release of National Security Agency documents demonstrated the consequences of the insider threat. More recently, the United States tried to attack North Korea but failed, South Korea was attempting to attack North Korea, and both applied Stuxnet-like approaches. These sophisticated targeted attacks differ from web-site hacking events that are reported almost daily in the news mainly because targeted attacks require detailed design and operation information of the systems attacked and/or are often carried out by insiders. For instance, in order to minimize disruption of facilities around the world, Stuxnet remained idle until it recognized the specific configuration of the Natanz facility, demonstrating that the attackers possessed extremely detailed information about the facility. Such targeted cyber attacks could become a national-level military weapon and be used in coercion of hostile countries. While U.S. nuclear power plants (NPPs) are well designed and protected and not easily attacked through internet hacking, there have been a few digital-system-related incidents. For example, in 2003, the Slammer worm breached the private network of the Davis Besse plant and disabled a safety monitoring system for almost five hours. The breach did not pose a safety hazard because the plant was offline, but it demonstrates the vulnerability of NPP systems. The worm did not enter the plant systems directly; it began by entering the systems of a Davis Besse contractor, and entered through the T1 line bridging the contractor's computer and Davis Besse's corporate networks. The T1 line was not protected by the plant's firewall. The success of the Stuxnet-Natanz case demonstrates that supply-chain level attack (i.e., physically tampering with digital systems to install undetectable malware for the purpose of bringing harm to a player further down the supply chain network) plays a crucial role in a successful attack. Although attacking an NPP naturally requires more complex strategies and tactics than internet hacking, NPP's have a variety of sophisticated and possibly sensitive systems that may be attacked. Previous studies, for example, Baylon, have identified and confirmed that such attack strategies and tactics can be designed and could produce severe damage to an NPP. Current NPPs were designed according to design basis accidents that predate use of the internet by the general public and did not take into consideration potential cyber attacks that could be carried out by state-sponsored attackers and malicious insiders. To date, no detailed engineering analysis has been performed to realistically examine the consequences these attacks may have on the plants. In this paper, we discuss an approach for assessing the potential consequences of Stuxnet type of attacks. We will examine the engineering criteria used in designing the plants, develop possible attack scenarios in terms of the structures, components, and systems (SSCs) being controlled by digital systems at the NPPs, perform engineering analysis to determine if the attacks would impose conditions that are beyond design bases, and categorize the consequences of the potential accidents. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Assessment of dynamic PRA techniques with industry average component performance data - 208

In the nuclear industry, risk monitors are intended to provide a point-in-time estimate of the system risk given the current plant configuration. Current risk monitors are limited in that they do not properly take into account the deteriorating states of plant equipment, which are unit-specific. Current approaches to computing risk monitors use probabilistic risk assessment (PRA) techniques, but the assessment is typically a snapshot in time. Living PRA models attempt to address limitations of traditional PRA models in a limited sense by including temporary changes in plant and system configurations. However, information on plant component condition is not considered. This often leaves risk monitors using living PRA models incapable of conducting evaluations with dynamic degradation scenarios evolving over time. There is a need to develop enabling approaches to solidify risk monitors to provide time- and condition-dependent risk by integrating traditional PRA models with condition monitoring and prognostic techniques. This paper presents an estimation of system risk evolution over time by integrating plant risk monitoring data with dynamic PRA methods that incorporate aging and degradation. Several online, nondestructive approaches have been developed for diagnosing plant component conditions in the nuclear industry, i.e., using a condition indication index, vibration analysis, current signatures, and operational history . In this work, the component performance measures at U.S. commercial nuclear power plants are incorporated within the dynamic PRA methodologies to provide better estimates of the probability of failures. Aging and degradation are modeled within the Level 1 PRA framework, are applied to several failure modes of pumps, and can be extended to a range of components, namely, valves, generators, batteries, and pipes. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

An overview of the IEEE human factors standard development activities - 2017 - 230

Since 1980 the Institute of Electrical and Electronics Engineers (IEEE) has supported development of human factors (HF) standards. Within IEEE, Subcommittee 5 (SC5) of the Nuclear Power Engineering Committee develops and maintains HF standards applicable to nuclear facilities. These standards are structured in a hierarchical fashion. The top-level standard defines the HF tasks required to support the integration of human performance into the design process. Five lower tier documents expand upon the upper tier standard. Presently, two new HF standards projects are underway; one to provide HF guidance for the validation of the system interface design and integrated systems operation and another for designing and developing computer-based displays for monitoring and control of nuclear facilities. In addition to producing and maintaining HF standards, SC5 is also involved in outreach activities, including sponsorship of a series of conferences on human factors and nuclear power plants. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Risk informed cyber security for nuclear power plants - 342

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

KGK-2-type detector of gamma-radiation power for diagnosis of nuclear reactor radiation fields within the range from 1 µGy/s to 100 Gy/s

The construction of the KGK-2-type detector of γ-radiation power is briefly described. The diagnostic possibilities of the detector are shown by the example of results of the dose rate measurement in the energy start-ups of the BR-K1 and BR-1M reactors implemented in the mode of generating fission pulses on delayed neutrons. The possibilities of using the KGK-2 detector for postpulse γ diagnostics are demonstrated by the example of results of measurements in the fission pulse on prompt neutrons of the BR-1M reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Implementation of wireless technologies in nuclear power plants' electromagnetic environment using cognitive radio system - 153

Wireless technologies have increased efficiency in many industrial settings across the world by enabling a mobile workforce and improving communications. Unfortunately, the nuclear power industry has been slow to adopt wireless technology as a result of safety, security, and reliability concerns expressed by regulators and others. A Research and Development (R and D) effort is currently being conducted to address two significant technical concerns for the introduction of wireless technology in nuclear power plants, which are electromagnetic compatibility (EMC) and wireless coexistence. Most equipment in nuclear power plants has never been tested for vulnerability to wireless transmission. As such, the impact of modern wireless devices on nuclear safety and plant reliability is not understood. Existing guidance from the Nuclear Regulatory Commission (NRC) and Electric Power Research Institute (EPRI) rely on the use of exclusion zones to protect plant equipment, which can result in exclusion zones of up to eight feet for some tablet devices rendering them ineffective for the mobile workforce. Another unknown is how different wireless devices will operate in close proximity to one another in the harsh electromagnetic environment of a nuclear power plant. These issues are being addressed through development of a cognitive radio system that has the ability to generate and output multiple wireless signals such as Wi-Fi, Bluetooth, and cellular communications at varying power levels and frequencies. The cognitive radio system is a modular system that can test nuclear plant equipment located in training areas, simulators, and the actual plant environment. In addition to using the system for testing plant equipment to wireless signals and EMC standards, the system can also monitor the radio spectrum for usage in other plant focuses such as cyber security. The R and D work will provide the objective guidance for the nuclear industry in the two areas mentioned and will develop a cognitive radio system for in-plant EMC and coexistence testing. The system will assist in the seamless use of wireless technology in nuclear power plants while greatly reducing the risks wireless signals may introduce. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Computerized operator support system and human performance in the control room - 262

An operator aid technology to assist nuclear plant operators maintain situation awareness and detect faults earlier than would be possible using conventional control room technologies is described. Coined the Computerized Operator Support System (COSS), this technology is aimed at the human-machine interface. The COSS has been implemented as a prototype system operating in the control room environment of a full-scale simulator for a commercial nuclear power plant. A human factors assessment that involved two licensed reactor crews was performed to evaluate crew performance. This paper describes the development of the COSS, the underlying fault-detection architecture, and feedback received based on operator-in-the-loop studies. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗