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At least 325 records · Page 18
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DOE OSTI2017

Eye tracking studies based on attentional-resource effectiveness - 310

In the majority of cases, the primary means of information input to operators in nuclear power plants (NPPs) control rooms is through the visual channel. In this study, eye movement patterns of NPP operators are analyzed with eye tracking data obtained from simulator-based experimental studies. Two eye tracking measures of attentional-resource effectiveness in monitoring and detection tasks in NPPs which had been developed by the authors are introduced and several applications with the two eye tracking measures are discussed for the use of the measures. The underlying principle of the measures is that information sources should be selectively attended according to their importance. One of the two measures is the fixation to importance ratio (FIR), which represents attentional resource (eye fixations) spent on an information source compared to the importance of the information source. The other measure is selective attention effectiveness (SAE), which incorporates the FIRs of all information sources. The FIR represents the effectiveness on an information source, whereas the SAE represents the overall effectiveness on all information sources. Frequency and the duration of eye fixations of an operator on information sources are used as attentional resource. Finally feasible future applications with eye tracking data were discussed in terms of plant systems automation and operator training. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Optimized application of regulatory human factors engineering guidelines - 314

In AREVA NP's recent projects of new plants or control room modernizations, the engineering of human factors was guided by NUREG 0711 and IEC 60964 and applied throughout the project phases. Each of the projects had a group of stakeholders, whose early mutual agreement on HFE matters was a prerequisite for successful HFE work. The group of main stakeholders comprises the utility's plant managers and members of the plant operating team, AREVA NP's project managers and members of the various technical competence centers, and representatives of the local safety authority. The objectives of each of these stakeholders and our advanced yet practice-proven HFE approach to successfully meet these objectives are described in this paper. It describes both our unique attention to the project's reference plant and its human-centered aspects thus inherited as well as our focus on those issues of the current plant which significantly depart from referenced practices. Deviations' significance and nature determine and grade the resulting HFE effort and its interdisciplinary conductance. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Test based reliability assessment method for a safety critical software in reactor protection system - 284

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

High-sensitivity fast neutron detector KNK-2-7M

The construction of the fast neutron detector KNK-2-7M is briefly described. The results of the study of the detector in the pulse-counting mode are given for the fissions of {sup 237}Np nuclei in the radiator of the neutron-sensitive section and in the current mode with the separation of sectional currents of functional sections. The possibilities of determining the effective number of {sup 237}Np nuclei in the radiator of the neutronsensitive section are considered. The diagnostic possibilities of the detector in the counting mode are shown by example of the analysis of the reference data from the neutron-field characteristics in the working hall of the BR-K1 reactor. The diagnostic possibilities of the detector in the current operating mode are shown by example of the results of measuring the {sup 237}Np-fission intensity in the BR-K1 reactor power start-ups implemented in the mode of fission-pulse generation on delayed neutrons at the detector arrangement inside the reactor core cavity under conditions of a wide variation of the reactor radiation field.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Experimental Design for Evaluating Selected Nondestructive Measurement Technologies - Advanced Reactor Technology Milestone: M3AT-16PN2301043

The harsh environments in advanced reactors (AdvRx) increase the possibility of degradation of safety-critical passive components, and therefore pose a particular challenge for deployment and extended operation of these concepts. Nondestructive evaluation technologies are an essential element for obtaining information on passive component condition in AdvRx, with the development of sensor technologies for nondestructively inspecting AdvRx passive components identified as a key need. Given the challenges posed by AdvRx environments and the potential needs for reducing the burden posed by periodic in-service inspection of hard-to-access and hard-to-replace components, a viable solution may be provided by online condition monitoring of components. This report identifies the key challenges that will need to be overcome for sensor development in this context, and documents an experimental plan for sensor development, test, and evaluation. The focus of initial research and development is on sodium fast reactors, with the eventual goal of the research being developing the necessary sensor technology, quantifying sensor survivability and long-term measurement reliability for nondestructively inspecting critical components. Materials for sensor development that are likely to withstand the harsh environments are described, along with a status on the fabrication of reference specimens, and the planned approach for design and evaluation of the sensor and measurement technology.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Flow Enhanced Electrochemical Sensor Performance in Complex Salt Systems

To advance MSR MC&A practices, Argonne National Laboratory developed and optimized robust flow enhanced electrochemical sensors (FEES) and multielectrode array voltammetry sensors (MAVS) which provide accurate near-real time measurements of actinides in molten salts. Flow-enhanced electrochemical sensors are installed directly into MSR flow conduits for inline salt chemistry measurement, while MAVS are deployed in quiescent salt conditions enabling online species concentration determination in stationary salt vessels. This report summarizes efforts to improve electrochemical sensor technological readiness through (1) sensor testing in complex, high concentration molten salt systems to demonstrate low uncertainty actinide measurements in MSR-representative solutions and (2) demonstrations of sensor performance in challenging real-world conditions in collaboration with partner MSR institutions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Multi-layered adaptative human-machine interface to support NPP operators - 104

The paper is aimed at improvement of situational awareness and reduction of information overload and during transients and abnormal operational modes. We investigate the possibility of implementation of a multi-layered interface concept for presentation of information to NPP operators. Four layers are suggested, namely: the structural layer representing process system in the form of a mimic diagram; the analytical layer intended to support cognitive activity performed by operator; the functional layer aimed at representation of the status of process functions performed by the plant equipment; and the reference layer containing any relevant information depending on the situation. All the layers can be superimposed on each other. Two factors influence on activation of layers, namely: category of operator and operational mode. The process of designing a multi-layered interface is based on functional analysis and illustrated in application to a simple heat generation system. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Study on the assessment method of human error probabilities in the digitalized main control room - 170

Recently, nuclear power plants (NPPs) in Republic of Korea have been constructed with the new type of main control room (MCR) which is called as digitalized MCR. In digitalized MCR, as digital technologies are being adopted, the environment of MCR has been considerably changed. The operators may obtain the plant data via digitalized human system interfaces (HSIs), large display panel (LDP), computerized procedure system (CPS), soft controls and other system features. Accordingly, the necessity to consider the new environment of MCR has been raised when performing human reliability analysis (HRA) methods. However, none of HRA methods consider the newly installed MCRs. The goal of this paper is to propose new frameworks to assess the human error probabilities (HEPs) in digitalized MCR. Generally, the HEPs are obtained by summing the probabilities of diagnosis errors and the probabilities of execution errors and those probabilities are modified by performance shaping factors (PSFs). In this paper, two parts are included; (1) Assessing the probabilities of diagnosis errors, and (2) Assessing the probabilities of execution errors. In this paper, the process to develop the frameworks are introduced and we expect that the HEPs affecting the features of digitalized MCR can be calculated by using the proposed frameworks. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Development of cyber security test scenario for non-safety display system - 226

In order to test the cyber security of NPP related to safety, the results of probabilistic safety assessment were used. The fault tree (FT) is used to reflect the impact of the cyber-attack on the operator. The cyber security test should be performed focusing on the scenario derived from the FT. The concept of a security case has been introduced. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Real-time fault detection and transient identification using extended Kalman filters - 254

In this paper, real-time fault detection and transient identification are studied with adaptive estimation methods, in particular parallel processing methods by implementing a bank of extended Kalman filters. Adaptive estimation is combined with probabilistic risk assessment to allow for real-time continuous monitoring of initiating events. The approach provides a probability of an initiating event happening. Real-time monitoring of initiating events can reduce labor demand. Each extended Kalman filter is configured to operate at a single configuration, either at normal operation or a faulted state. Fault detection is based on calculating a likelihood function for each extended Kalman filter to determine the posteriori conditional probability that the system configuration is the true configuration. The transient identification is determined by performing a multivariate statistical analysis on each extended Kalman filter to provide a posterior probability assessment that the reactor states and its dynamics are within the allowable limits. The probability distribution being outside the allowable limits indicate a transient has occurred. The combination of parallel processing methods with the multivariate statistical analysis for transient identification allows for the operator to perform transient identification with having a high level of certainty of the current system configuration. We demonstrate the real-time fault detection and transient identification on a simple reactor example to determine the reactor power, reactor temperature, and coolant temperature in the presence of various faults. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Technical constraints to conducting operator studies in a reconfigurable control room simulator - 269

The Human Systems Simulation Laboratory (HSSL) at Idaho National Laboratory (INL) is a full scope, full scale, and reconfigurable digital representation of an analog nuclear power plant control room. The HSSL is used to conduct human factors research studies on digital technologies being introduced into legacy nuclear power plant control rooms. With the aging nuclear fleet approaching an average of 35 years, utilities are beginning to replace their analog systems with more modern technologies, and utilizing the HSSL to accomplish these tasks. This document will describe some of the constraints identified when conducting operator studies in the HSSL. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

A human factors meta model for U.S. nuclear power plant control room modernization - 350

Over the last several years, the United States (U.S.) Department of Energy (DOE) has sponsored human factors research and development (R and D) and human factors engineering (HFE) activities through its Light Water Reactor Sustainability (LWRS) program to modernize the main control rooms (MCR) of commercial nuclear power plants (NPP). Idaho National Laboratory (INL), in partnership with numerous commercial nuclear utilities, has conducted some of this R and D to enable the life extension of NPPs (i.e., provide the technical basis for the long-term reliability, productivity, safety, and security of U.S. NPPs). From these activities performed to date, a human factors meta model for U.S. NPP control room modernization can now be formulated. This paper discusses this emergent HFE meta model for NPP control room modernization, with the goal of providing an integrated high-level road map and guidance on how to perform human factors R and D and HFE for those in the U.S. nuclear industry engaged in the process of upgrading their MCRs. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Design themes for future hybrid nuclear power plant control rooms - 383

As nuclear power plants (NPP) control rooms are updated and new technologies added, a new hybrid control room is set to emerge and with it new safety implications for design. This paper highlights two technologies determined to be highly relevant for control room modernization: overview displays and computer-based procedures. Control room operators were observed in the simulator, while conducting abnormal and emergency scenarios. Additionally, operators were observed and interviewed in the control room during normal operations. Design themes for both of the technologies are discussed. Results revealed ten main design themes for overview displays, including observations of the system used currently in order to provide overview information. The results revealed three main design themes for computer-based procedures that were more directed at an envisioned world, since these technologies are not deployed currently. This evaluation elicited considerations for requirements for future control room system designs and recommendations for hybrid NPP control rooms. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Review Paper: Review of Instrumentation for Irradiation Testing of Nuclear Fuels and Materials

Over 50 years of nuclear fuels and materials irradiation testing has led to many countries developing significant improvements in instrumentation to monitor physical parameters and to control the test conditions in material test reactors (MTRs). Recently, there is increased interest to irradiate new materials and reactor fuels for advanced pressurized water reactors and Gen-IV reactor systems, such as sodium-cooled fast reactors, very high temperature reactors, supercritical water-cooled reactors, and gas-cooled fast reactors. This review paper documents the current state of instrumentation technologies in MTRs in the world and summarizes ongoing research efforts to deploy new sensors. As described in this paper, a wide range of sensors is available to measure key parameters of interest during fuels and materials irradiations in MTRs. Ongoing development efforts focus on providing MTR users a wider range of parameter measurements with smaller, higher accuracy sensors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Achieving reasonable confidence in the validation of control room designs and modifications: A summary of the 2015 NEA experts' workshop - 124

Defining and achieving reasonable confidence in the validation of nuclear power plant control room designs is a complex challenge for many reasons; not the least of which is that validation is a test of strength rather than a test of truth. In 2015, the Nuclear Energy Agency, Committee for Safety of Nuclear Installations' Working Group on Human and Organizational Factors (WGHOF) hosted an experts' workshop to identify and critically examine potential means to address the challenge of achieving reasonable confidence in control room design validations. The experts' principal task was to answer four challenge questions. (1) What are the critical considerations in defining validation objectives and how do these impact achieving reasonable confidence? (2) What methods, approaches, resources, or rationales might be used for deriving performance requirements, selecting measures, and establishing acceptance criteria so as to support reasonable confidence? (3) What methods might be used to develop scenarios that maximize the amount and relevance of information in support of the validation conclusions and achieving reasonable confidence? (4) How should the validation results be aggregated and analyzed to determine the final validation conclusions? This paper provides an overview of the experts' responses to these challenges and briefly touches upon their implications for future directions. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗