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DOE OSTI2017

Identifying requisite changes to support modern maintenance practices - 294

Modern data science and processing require the use of modern maintenance records. With traditional record-keeping practices, the isolated nature of the data, as well as the presence of handwritten maintenance work orders and incomplete service logs, severely limit the amount of progress that can be made towards advanced condition-based maintenance. These condition-based methods of diagnostics, prognostics, and maintenance planning are a primary goal of next-generation plant operations and maintenance strategies, and are all dependent on the availability of quality plant data including maintenance data. By implementing changes to maintenance record-keeping practices, such as the use of a computerized maintenance management system, key data will become available to allow for implementation of state-of-the-art data analysis methods. This paper covers the benefits of utilizing digital record systems to increase maintenance data availability in terms of condition-based maintenance and asset management. The ability of these methods to enable state-of-the-art practices within industrial organizations is discussed to offer insight into ways that these improvements to operating strategy will contribute to improvements in world-class maintenance excellence. A discussion of the future of asset management in terms of data analytics, maintenance decision support, and data integration is provided to illustrate the changing landscape of plant management. The result is an examination of innovative maintenance strategies and discussion of challenges related to the future of asset management and implementing modern maintenance practices. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2016

A Study on Safety Evaluation of Safety Grade Smart Transmitter in Nuclear Power Plants

Smart transmitters contain microprocessors and the use of firmware and software in these microprocessors presents challenges to the nuclear industry, particularly when they are used in safety systems. Non-safety grade smart transmitters have been used for I and C systems of Nuclear Power Plants (NPPs). Recently, smart transmitters have been used for safety grade I and C systems as well as non-safety grade I and C system for SKN 3 and 4 NPPs in Korea. Due to potential benefits of smart transmitter, it is anticipated smart transmitters will be widely used safety-related applications in NPPs. But there is currently no specific nuclear sector standard specifying the criteria for smart transmitter. Therefore, Smart transmitter's technology and characteristics need to be investigated and also develop review guideline of smart transmitter. For this reason, we performed a study on safety evaluation of safety grade Smart Transmitter in NPPs. This paper presents the results of study on safety evaluation of smart transmitter. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

Assessment of Flow-Enhanced Electrochemical Sensor Testing and Deployments for MSRs

This report serves as the deliverable for Milestone M3RS-23AN0401061 that is part of Work Package RS-23AN040106 (Flow Enhanced Sensors for MSRs – ANL). The goal of this milestone was to determine performance of the flow enhanced electrochemical sensor (FEES) and modular flow instrumentation testbed (MFIT) in safeguards relevant scenarios. Flow enhanced electrochemical sensors are a type of electroanalytical sensor that has been developed at Argonne National Laboratory to be installed directly into MSR flow conduits to make measurements of the salt composition. These sensors represent a significant improvement in capabilities compared to earlier electroanalytical sensors that instead can only be operated in quiescent conditions. Previous work has focused on testing of the FEES in flowing conditions provided by the MFIT to assess the accuracy and precision of the sensor measurements. To further improve this capability, in FY23 we undertook a campaign of safeguards relevant scenarios in molten salt containing a range of uranium chloride concentrations (0 to 3 wt%). All the testing carried out in FY23 was aided by a control system designed to automatically actuate flow conditions and collect data. This new automation system is estimated to have increased experimental throughput by a factor of four and enabled testing in a variety of complex conditions. The advancements in throughput and repeatability led to improved quantification of actinide concentrations using the in-flow sensors, with a reduction of the mean absolute relative error from 5.6% in FY22 to 3.1% in FY23. In addition to safeguards scenarios run in the MFIT, FY23 work included deployment of a FEES at a partner institution where it will be tested in a flowing salt loop. The FEES was successfully integrated into that loop and is being tested prior to loop startup. In FY23, work also continued on the smaller flow system that we have named the mini-MFIT. This smaller system is capable of rapid prototyping of new sensor designs prior to installation in the larger MFIT radiological flow system. Work was carried out to test this new system in non-radiological molten salts in a separate glovebox. This work is helping us to enhance the accuracy of our salt monitoring capabilities through the integration of multiple types of sensors. The high degree of accuracy required by 10 CFR 74 represents a significant challenge, and further design evolution and integration of the sensors into multimodal sensing frameworks will be needed to further push the measurement accuracy to the needed level.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Analysis of SPND Material Candidates for Real-Time Neutron Spectrum Monitoring Poster

The future of the nuclear industry will be shaped by advanced reactor designs requiring innovative instrumentation for safe operation. A key area of innovation is the ability to accurately and reliably monitor the neutron energy spectrum during operations, which would also enhance the experimental capabilities of existing research reactors. To address this need, our research focuses on the potential use of Self-Powered Neutron Detectors (SPNDs) for real-time neutron spectrum monitoring. We investigated various candidate materials for SPNDs by evaluating their energy-dependent reaction rates. Promising materials were further modeled using MCNP to determine the free electron production per neutron across the neutron energy spectrum, approximating their energy-dependent signal sensitivity. Our findings identified Samarium (Sm) and Gadolinium (Gd) as top candidates for thermal neutrons, Hafnium (Hf) and Thulium (Tm) for epithermal neutrons, and Tantalum (Ta) for fast neutrons. These materials will undergo future experimental validation to confirm the modeled responses and assess their feasibility for real-time neutron spectrum measurements.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗

DOE OSTI1995

Nuclear Safety [Vol. 36, No. 2, July-December 1995]

Nuclear Safety is a journal that covers significant issues in the field of nuclear safety. Its primary scope is safety in the design, construction, operation, and decommissioning of nuclear power reactors worldwide and the research and analysis activities that promote this goal, but it also encompasses the safety aspects of the entire nuclear fuel cycle, including fuel fabrication, spent-fuel processing and handling, and nuclear waste disposal, the handling of fissionable materials and radioisotopes, and the environmental effects of all these activities. Table of Contents for this issue follows. THE CHORNOBYL ACCIDENT: 195 The Chornobyl Accident Revisited, Part III: Chernobyl Source Term Release Dynamics and Reconstruction of Events During the Active Phase, A. R. Sich; GENERAL SAFETY CONSIDERATIONS: 218 Second ANS Workshop on the Safety of Soviet-Designed Nuclear Power Plants, R. A. Bari; 234 Elements of a Nuclear Criticality Safety Program, C. M. Hopper; 243 Rickover, Excellence, and Criticality Safety Programs, R. E. Wilson; ACCIDENT ANALYSIS: 249 Transient Analysis of the PIUS Advanced Reactor Design with the TRAC-PF1/MOD2 Code, B. E. Boyack, J. L Steiner, S. C. Harmony, H. J. Stumpf, and J. F. Lime; 278 The Hierarchy-By-Interval Approach to Identifying Important Models that Need Improvement in Severe-Accident Simulation Codes, T. J. Heames, M. Khatib-Rahbar, J. E. Kelly, R. P. Jenks-Johnson, and Y.-S. Chen; 290 RELAP5/MOD3 Code Coupling Model, R. P. Martin; 299 Missiles Caused by Severe Pressurized-Water Reactor Accidents, R. Krieg; DESIGN FEATURES: 310 Validation of COMMIX with Westinghouse AP-600 PCCS Test Data, J. G. Sun, T. H. Chien, J. Ding, and W T. Sha; ENVIRONMENTAL EFFECTS: 321 Spent Nuclear Fuel Characterization for a Bounding Reference Assembly for the Receiving Basin for Off-Site Fuel, S. D. Kahook, R. L. Garrett, L. R. Canas, and M J. Beckum; OPERATING EXPERIENCES: 332 Reactor Shutdown Experience, Compiled by J. W. Cletcher; U.S. NUCLEAR REGULATORY COMMISSION INFORMATION AND ANALYSES: 335 Reactor Coolant System Blowdown at Wolf Creek on September 17, 1994, J. V. Kauffman and S. L. Israel; RECENT DEVELOPMENTS: 344 Reports, Standards, and Safety Guides, D. S. Oueener; 349 Proposed Rule Changes as of June 30, 1995; ANNOUNCEMENTS: 320 Symposium on Acceptability of Risk From Radiation—Application to Manned Space Flight; 320 24th DOE/NRC Nuclear Air Cleaning and Treatment Conference; 361 Radiation Biology and Radiation Protection— Modern Developments and Tendencies in Radiation Biology; 362 1997 IEEE Sixth Conference on Human Factors and Power Plants; 354 The Authors; 360 Reviewers of Nuclear Safety, Vol. 36.

05 NUCLEAR FUELS↗

DOE OSTI1996

Nuclear Safety [Vol. 37, No. 2, April-June 1996]

05 NUCLEAR FUELS↗

DOE OSTI2017

Effects of adaptive automation on operator-system performance - 103

Adaptive automation (AA) is the dynamic, real-time change in the degree of automation in response to situational changes. AA systems are being developed in many industries, including for nuclear power plant applications. AA is a promising means of mitigating human performance issues that often arise in highly automated systems. The purpose of this Nuclear Regulatory Commission study was to characterize important AA design features, define the effects of AA on performance, examine the human factors engineering (HFE) guidance available to design and evaluate AA systems, and identify research and development needs. In general, we found that AA improves task performance and situation awareness (SA) while managing workload. Although the research is limited, AA also supports operator recognition of automation failure and recovery. Research on human-system interfaces has identified principles showing promise for supporting SA, including failure detection, and limiting workload. Research further suggests that AA systems are both more effective, and less disruptive and distracting, when they follow rules of interaction similar to those used by human crewmembers. Our general conclusion about the HFE guidance available to designers and reviewers is that it is limited and has not kept pace with automation technology. We discuss several research and development needs that will help support a better understanding of where and how AA should be used and support the development of more comprehensive guidance for AA safety reviews. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Supporting resilience in nuclear power plant operations - conceptual frames for human factors development and validation - 166

This is a summary paper of conceptual developments on nuclear power plant operation by the human factors team at VTT Technical Research Centre of Finland. While drawing together years of studies, the paper also discusses recent progresses in training development. Firstly, there is a long background in work analysis and system validation. The analysis model applied provides a model for assessing and identifying 'core-task functions,' that is, basic functions that should be fulfilled within work activity. Secondly, we have evaluated the existing training practices at a Finnish plant and identified challenges that could be generally prevalent in the nuclear domain - they seem to reflect the hierarchical and safety-critical nature of NPP operation. In terms of potential solutions, we are currently testing and developing some new self-evaluation and self-confrontation methods. First and foremost, however, we discuss the common theory basis for our research approaches. On the broadest level, our development and research work reflects the view of safety as expressed by resilience engineering literature: safety is not seen merely as 'negative' lack of mistakes, but also as 'positive' capability to solve and anticipate problems. For example, it is assumed that operators' capability for situational interpretation should be supported at all times: even if the emergency procedures dictate operator activity specifically, the operators should understand the influence of the operating procedures for the plant process and system state. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Development of a model based assessment process for qualification of embedded digital devices in NPP applications: research approach and current status - 263

The instrumentation and control (I and C) equipment used in currently operating U.S. nuclear power plants (NPPs) is primarily based on mature analog technologies that are progressing towards obsolescence. The continued reliance on this legacy analog technology, which is also being propagated into new NPP designs, imposes performance penalties and maintenance burdens in comparison with modern digital I and C instrumentation. In many instances, currently available I and C equipment contain embedded digital devices (EDDs) such as microprocessors and programmable logic devices. Experience in other industries, such as avionics, has shown that digital I and C equipment containing EDDs can provide significant benefits over analog-based equipment in terms of performance, reliability, and maintainability. In recent years, due to the high demand for digital technologies in the industrial I and C marketplace, it is becoming increasingly difficult for NPPs to acquire instrumentation that is not equipped with an EDD. However, the nuclear power industry has been slow to adopt digital technology, especially in safety-related systems, in large part as a result of regulatory concerns about common-cause failure (CCF) vulnerabilities of equipment with EDDs. Consequently, there is a clear need to develop cost effective qualification methods to contribute to the assessment of CCF vulnerability posed by EDDs in modern instrumentation that could be used in NPPs. This paper describes the research regarding qualification methods for equipment with EDDs that is sponsored by the Nuclear Energy Enabling Technologies (NEET) Advanced Sensors and Instrumentation (ASI) program of the U.S. Department of Energy (DOE). The purpose of the current research is to develop an effective approach employing science-based methods to resolve concerns about CCF vulnerability that serve to inhibit deployment of advanced instrumentation (e.g., sensors, actuators, microcontrollers) with EDDs in nuclear power applications. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Development of the Delphi display concept for alternative explanations decision making support - 120

Human operators of advanced technology systems, such as nuclear power, command and control, and deepwater oil drilling, must be able to evaluate and respond to a wide array of operational anomalies. Unfortunately, these circumstances are often complex and ambiguous, complicating decision making tasks. One common decision making error, called confirmation bias, occurs when decision makers prematurely focus on one explanation instead of systematically considering all plausible alternative explanations that could equally well account for an operational anomaly (Nickerson, 1998). The Delphi display is designed to counteract decision makers' tendencies towards confirmation bias and to encourage them to consider all plausible explanations, including high risk explanations, they might otherwise overlook. An initial Delphi display prototype was designed to help identify the underlying cause of operational anomalies in the complex, high-consequence technology of deepwater drilling. This prototype decision aid incorporates major indicators related to operational anomalies and presents corresponding plausible explanations for each indicator, ranging from high-risk system-critical explanations to lower-risk but still significant events. The Delphi display incorporates several innovative features to help decision makers detect and monitor operational anomalies and view the relationships among indicators and their associated plausible explanations. Preliminary evaluations of an early prototype Delphi display for detecting kicks in deepwater oil drilling operations have been promising. Moreover, the methodology used to design and develop the Delphi display can be readily generalized to numerous other operational contexts that require humans to make system-critical decisions. To illustrate, an initial Delphi display design for nuclear power plant control is described. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

Development of utility generic functional requirements for electronic work packages and computer-based procedures - 273

The Nuclear Electronic Work Packages - Enterprise Requirements (NEWPER) initiative is a step toward a vision of implementing an eWP framework that includes many types of eWPs. This will enable immediate paper-related cost savings in work management and provide a path to future labor efficiency gains through enhanced integration and process improvement in support of the Nuclear Promise (Nuclear Energy Institute 2016). The NEWPER initiative was organized by the Nuclear Information Technology Strategic Leadership (NITSL) group, which is an organization that brings together leaders from the nuclear utility industry and regulatory agencies to address issues involved with information technology used in nuclear-power utilities. NITSL strives to maintain awareness of industry information technology-related initiatives and events and communicates those events to its membership. NITSL and Light Water Reactor Sustainability Program researchers have been coordinating activities, including joint organization of NEWPER-related meetings and report development. The main goal of the NEWPER initiative was to develop a set of utility generic functional requirements for eWP systems. This set of requirements will support each utility in their process of identifying plant-specific functional and non-functional requirements. The NEWPER initiative has 140 members where the largest group of members consists of 19 commercial U.S. nuclear utilities and eleven of the most prominent vendors of eWP solutions. Through the NEWPER initiative two sets of functional requirements were developed; functional requirements for electronic work packages and functional requirements for computer-based procedures. This paper will describe the development process as well as a summary of the requirements. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

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

Instrumentation and control strategies for an integral pressurized water reactor

Not Available

Upadhyaya, Belle R.↗

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↗