Search NASA⌕ Search

SEARCH · Search NASA

Results for “REACTOR INSTRUMENTATION”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 361 records · Page 20
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Early implementation of digital upgrades on simulators - 347

An important aspect of successful Nuclear Power Plant (NPP) digital control system upgrade projects is to upgrade the Operator Training Simulator (OTS) or a related Engineering Simulator (ES), prior to installation of the Distributed Control System (DCS) on the plant allowing for early testing of the digital controls/HMI and for early training of the plant operators while maintaining OTS configuration management (i.e. OTS in sync with the plant's current design). Experience has shown that early implementation of digital upgrades on the OTS and/or an ES can reduce plant startup duration (related to digital systems commissioning) and ensure operators are proficient with the new digital systems to ensure safe and reliable plant operation. Including simulators in the design process offers many benefits including early identification of design deficiencies, testing deficiency fixes, developing and validating plant procedures and supporting human factors engineering, to name a few. Implementing digital upgrades on the OTS is also a regulatory requirement driven by the ANSI/ANS-3.5-2009 standard (Nuclear Power Plant Simulators for Use in Operator Training and Examination) if the new DCS affects actions to be taken by the plant operator in the main control room. This paper presents the benefits of implementing digital upgrades on the OTS and/or ES prior to plant implementation. The various simulator implementation strategies (stimulation, emulation, simulation or a hybrid solution) and selection factors are discussed. The paper also covers four (4) case studies demonstrating how digital upgrades can be implemented on simulators. The first case study covers the implementation of a Siemens offgas Programmable Logic Controller (PLC) system on the Enrico Fermi OTS (Michigan). The second case study covers the implementation of a GE main generator excitation system on the Cernavoda OTS (Romania). The third case study covers the implementation of a Siemens main feedwater control system on the Callaway OTS and ES (Missouri). The final case study covers the implementation of the non-safety related and safety related DCS on the Nuclear New Build (NNB) Hongyanhe Phase I dual-purpose OTS/ES (China). (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Failure Mechanism Traceability and Application in Human System Interface of Nuclear Power Plants using RESHA

In recent years, there has been considerable effort to modernize existing and new nuclear power plants with digital instrumentation and control systems (DI&C). However, there has also been considerable concern both by industry and regulatory bodies for the risk and consequence analysis of these systems. Of particular concern are digital common cause failures (CCFs) specifically related to software defects. These “misbehaviors” by the software can occur in both the control and monitoring of a system. While many new methods have been proposed to identify potential software failure modes, such as Systems-theoretic Process Analysis (STPA), Hazard and Consequence Analysis for Digital Systems (HAZCADS), etc., these methods are focused primarily on the control action pathway of a system. In contrast, the information feedback pathway lacks unsafe control actions (UCAs), which are typically related to software basic events; thus, assessment of software basic events in such systems is unclear. In this work, we present the idea of intermediate processors and unsafe information flow (UIF) to help safety analysts trace failure mechanisms in the feedback pathway and how they can be integrated into a fault tree for improved assessment capability. The concepts presented are demonstrated in two comprehensive case studies, a smart sensor integrated platform for unmanned autonomous vehicles and another on a representative advanced human system interface (HSI) for safety critical plant monitoring. The qualitative software basic events are identified, and a fault tree analysis is conducted based on a modified Redundancy-guided Systems-theoretic Hazard Analysis (RESHA) methodology. The case studies demonstrate the use of UIF and intermediate processors in the fault tree to improve traceability of software failures in highly complex digital instrumentation feedback. The improved method can also clarify fault tree construction when multiple component dependencies are present in the system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Failure Mechanism Traceability and Application in Human System Interface of Nuclear Power Plants using RESHA

In recent years, there has been considerable effort to modernize existing and new nuclear power plants with digital instrumentation and control systems (DI&C). However, there has also been considerable concern both by industry and regulatory bodies for the risk and consequence analysis of these systems. Of particular concern are digital common cause failures (CCFs) specifically related to software defects. These “misbehaviors” by the software can occur in both the control and monitoring of a system. While many new methods have been proposed to identify potential software failure modes, such as Systems-theoretic Process Analysis (STPA), Hazard and Consequence Analysis for Digital Systems (HAZCADS), etc., these methods are focused primarily on the control action pathway of a system. In contrast, the information feedback pathway lacks unsafe control actions (UCAs), which are typically related to software basic events; thus, assessment of software basic events in such systems is unclear. In this work, we present the idea of intermediate processors and unsafe information flow (UIF) to help safety analysts trace failure mechanisms in the feedback pathway and how they can be integrated into a fault tree for improved assessment capability. The concepts presented are demonstrated in two comprehensive case studies, a smart sensor integrated platform for unmanned autonomous vehicles and another on a representative advanced human system interface (HSI) for safety critical plant monitoring. The qualitative software basic events are identified, and a fault tree analysis is conducted based on a modified Redundancy-guided Systems-theoretic Hazard Analysis (RESHA) methodology. The case studies demonstrate the use of UIF and intermediate processors in the fault tree to improve traceability of software failures in highly complex digital instrumentation feedback. The improved method can also clarify fault tree construction when multiple component dependencies are present in the system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Radiation hardened successive-approximation ADC with error detection circuits - 238

For several decades, radiation-hardened-by-design (RHBD) techniques have been developed to meet the design requirements of irradiating environment in nuclear power plants. Improvements on the circuit side for radiation sensors in performance, chip size, and radiation hardening ability have been adopted in current plant systems; however, next generation reactors and/or preparation for severe events in existing reactors require advanced circuit structures that can provide relatively long viability in harsh conditions. In order to maintain performances of electronics in high radiating environments, we propose to develop a small-area, low resource-overhead data-converter architecture that is immune to radiation impact events, an architecture that includes three kinds of novel radiation-hardened (rad-hard) designs for each analog and digital circuit in an ADC structure: a) error detectable tri-state buffer for flip flop against DSET, b) error detection storage in flip flop against SEU, and c) analog algorithm for the analog parts of an SAR ADC. The novel ADC architecture including the three RHBD techniques was implemented in a standard 180 nm CMOS technology with a 1.8 V supply voltage. Simulation results show that the techniques can successfully recognize errors induced by radiation impact events. The static power consumption was 123 {mu}W with the sampling rate 36 MS/s and resolution of 8 bits. The performances could be highly advanced under fine-tune designs. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI1991

Nuclear Safety [Vol. 32, No. 4, October-December 1991]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 477 Report on the International Symposium on the Use of Probabilistic Safety Assessment for Operational Safety—PSA '91, S. Chakraborty and M. Khatib-Rahbar; 488 Good Relationships Are Pivotal in Nuclear Data Bases, A. S. Heger and B. V. Koen; 494 Technical Note: The Interagency Nuclear Safety Review Panel's Evaluation of the Ulysses Space Mission, J. A. Sholtis, Jr., D. A. Huff, L. B. Gray, N. P. Klug, and R. O. Winchester; ACCIDENT ANALYSIS: 502 The Severe Accident Analysis Program for the Savannah River Nuclear Production Reactors, M. L. Hyder; CONTROL AND INSTRUMENTATION: 511 A Framework for Selecting Suitable Control Technologies for Nuclear Power Plant Systems, R. A. Kisner; DESIGN FEATURES: 521 Containments for Gas-Cooled Power Reactors History and Status, P.W. Williams; ENVIRONMENTAL EFFECTS: 537 Indoor Radon: A Natural Risk, N. H. Harley and J.H. Harley; On the Importance of the Atmospheric Parameters in the Fission Products Distribution of a Severe Reactor Accident, M. C. Barla and A. R. Bayulken; Book Review: Health Effects of Exposure to Low Levels of Ionizing Radiation BEIR V, C. R. Richmond; WASTE AND SPENT FUEL MANAGEMENT: 555 Activities Related to Waste and Spent Fuel Management, M. D. Muhlheim and E. G. Silver OPERATING EXPERIENCES: 567 Effects of Component Aging on the Westhinghouse Control Rod Drive System, K. Sullivan and W. Gunther; 577 Reactor Shutdown Experience, Compiled by J. W. Cletcher, 580 Selected Safety-Related Events, Compiled by G. A. Murphy; 582 Operating U 8 Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver, RECENT DEVELOPMENTS: 596 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 610 Reports, Standards, and Safety Guides, D. S. Queener; 614 Proposed Rule Changes as of June 30, 1991; ANNOUNCEMENTS: 520 Workshop on PC PRAISE: A Probabilistic Fracture Mechanics Personal Computer Code for Nuclear Power Plant Piping Reliability Assessment; 536 Fifth Workshop on Nuclear Power Plant Containment Integrity; 624 New OECD "International Information System on Occupational Exposure (ISOE)"; 625 Harvard School of Public Health Announces Short Courses; 625 Eighth Power Plant Dynamics, Control and Testing Symposium; 626 Second Training Course on Off-Site Emergency Planning and Response for Nuclear Accidents; 618 The Authors; 622 Reviewers of Nuclear Safety, Vol 32.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Criticality detection method based on FP gamma radiation measurement

A method is proposed to evaluate the extent of subcriticality of an accident-damaged nuclear reactor. In this method the radiation ratio of two FP rare gas nuclides, {sup 88}Kr and {sup 133}Xe, is measured. From the measured value, the value in the core is estimated by correcting for the time required by the gases to diffuse from the core to the measuring point. A simple expression for neutron multiplication factor has been derived, which uses the corrected {sup 88}Kr-to-{sup 133}Xe radiation ratio and the {sup 88}Kr-to-{sup 133}Xe fission yield ratios from {sup 244}Cm and {sup 235}U but no information on the amount or distribution of fissile materials, making the proposed method very simple. The method has the advantage that FP rare gases can easily leak from the reactor core through many openings and gaps, reaching germanium counters without reacting with other materials. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Business Case Analysis for Digital Safety-Related Instrumentation & Control System Modernizations

The commercial nuclear sector faces unprecedented financial challenges driven by low natural gas prices and subsidized renewables in a marketplace that does not reward carbon-free baseload capacity. These circumstances, along with increasingly antiquated labor-centric operating models and analog technology, have forced the premature closure of multiple nuclear facilities and placed a much larger population of nuclear power stations at risk. Nuclear plant economic survival in current and forecasted market conditions requires an efficient and technology-centric operating model that harvests the native efficiencies of advanced technology. This is analogous to transformations in nearly every other industry. In light of previous industry experience in modernizing safety Instrumentation and Controls (I&C) systems, nuclear utilities are reluctant to pursue these upgrades due to uncertainty in licensing and cost.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Human Factors Principles in Information Dashboard Design - 278

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)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Our Next Two Steps for Fukushima Daiichi Muon Tomography

After the vast disasters caused by the great earthquake and tsunami in eastern Japan, we proposed applying our Muon Tomography (MT) technique to help and improve the emergency situation at Fukushima Daiichi using cosmic-ray muons. A reactor-tomography team was formed at LANL which was supported by the Laboratory as a response to a request by the former Japanese Prime Minister, Naoto Kan. Our goal is to help the Japanese people and support remediation of the reactors. At LANL, we have carried out a proof-of-principle technical demonstration and simulation studies that established the feasibility of MT to image a reactor core. This proposal covers the next two critical steps for Fukushima Daiichi Muon Imaging: (1) undertake case study mock-up experiments of Fukushima Daiichi, and (2) system optimization. We requested funding to the US and Japanese government to assess damage of reactors at Fukushima Daiichi. The two steps will bring our project to the 'ready-to-go' level.

47 OTHER INSTRUMENTATION↗
DOE OSTI2017

Addressing embedded digital devices in safety-related systems of nuclear power plants - 307

The purpose of this paper is to provide an overview of the development of the approach to addressing embedded digital devices (EDD) in safety-related systems of nuclear power plants in the U.S. and around the world. In April, 2016, the U.S. Nuclear Regulatory Commission (US NRC) issued RIS 2016-05 to heighten awareness that EDDs might exist in procured equipment used in safety-related systems without the devices having been explicitly identified in procurement documentation. Inadequate consideration of these devices in digital technology system upgrades, component replacements, and new equipment applications could lead to an adverse safety consequence. The US NRC request was for addressees to implement early efforts to identify these devices. An area where this is being addressed is in regard to the performance of Diversity and Defense-In-Depth (D3) analyses, as addressed in US NRC Branch Technical Position 7-19. Nuclear facilities have increased their use and reliance on digital technology in systems and equipment (e.g., I and C, electrical systems, and fluid systems). In addition to digital instrumentation and control (I and C) systems, examples of safety-related equipment that may use digital technology include emergency diesel generators, pumps, valve actuators, motor control centers, breakers, priority logic modules, time-delay relays, and uninterruptible power sources. In the U.S. and around the world, engineering and licensing activities in standards and guidance have been, and are being developed to address this important consideration in protecting safety-related systems. This paper addresses development of an extended approach to evaluating common-cause failure vulnerability and mitigation in regard to equipment with an EDD. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Recent Developments and Findings of Heat Pipe Experiments for Microreactor Applications

Microreactor technologies are required to provide reliable carbon-free power generation in remote applications. The heat pipe–cooled microreactor concept, in particular, offers notable advantages due to the passive operation of heat pipes, enabling increased reliability and simplicity in a more compact form factor. There is a significant need for experimental work to aid and expedite the deployment of heat pipe microreactors due to their unique technological characteristics. Thus, there has been increased interest in heat pipe experiments by numerous institutions in order to support these efforts. Finally, the present work is a comprehensive review of recent heat pipe experiments from six major institutions, describing their designs, instruments, methods, and results. In addition, this paper also presents a background on heat pipe experiments along with discussions on instrumentation, accident scenarios, wick enhancement, and proposed future directions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Westinghouse Small Modular Reactor balance of plant and supporting systems design

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the second in a series of four papers which describe the design and functionality of the Westinghouse SMR. It focuses, in particular, upon the supporting systems and the balance of plant (BOP) designs of the Westinghouse SMR. Several Westinghouse SMR systems are classified as safety, and are critical to the safe operation of the Westinghouse SMR. These include the protection and monitoring system (PMS), the passive core cooling system (PXS), and the spent fuel cooling system (SFS) including pools, valves, and piping. The Westinghouse SMR safety related systems include the instrumentation and controls (I and C) as well as redundant and physically separated safety trains with batteries, electrical systems, and switch gears. Several other incorporated systems are non-safety related, but provide functions for plant operations including defense-in-depth functions. These include the chemical volume control system (CVS), heating, ventilation and cooling (HVAC) systems, component cooling water system (CCS), normal residual heat removal system (RNS) and service water system (SWS). The integrated performance of the safety-related and non-safety related systems ensures the safe and efficient operation of the Westinghouse SMR through various conditions and transients. The turbine island consists of the turbine, electric generator, feedwater and steam systems, moisture separation systems, and the condensers. The BOP is designed to minimize assembly time, shipping challenges, and on-site testing requirements for all structures, systems, and components. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Development of programmable logic in a rapid, automated, and repeatable fashion for safety-related applications - 385

Digital safety platform considerations for power plant modernization or new build efforts involve a burdensome task of implementing a large number of software-based functions onto a designated platform to match the plant's architecture. This problem is not unique to any particular safety-related platform solution, but is prevalent across the landscape of current Qualified and Licensed Safety-related Instrumentation and Control (I and C) Solutions. As an example when implementing a RG 1.97 compliant Post Accident Monitoring System (PAMS) solution with the Lockheed Martin Nuclear Protection and Control (NuPAC) safety platform, 72 unique software loads are needed to configure the Qualified Display System (QDS). Large scale development and customization efforts to upgrade existing plant functions or systems require the licensee to assume the cost and technical risk necessary to undergo a digital based upgrade. Often the cost, schedule, and technical risk outweighed the benefit of modernization efforts due to the unknown hurdles and past projects that overran cost and schedule projections. Lockheed Martin's Engineering Production Suite (EPS) minimizes the licensee risk and expense associated with modernization efforts through a developer's toolkit for rapid, automated, and repeatable development of Programmable Logic used to implement plant functions and architectures for use on the Lockheed Martin NuPAC platform. This paper describes a use case where by EPS can create system level representations, simulate system functions, partition logic designs to accommodate hardware constraints through logic partitioning, automate the assembly of Application Specific Programmable Logic (ASPL), generate system test cases, program the NuPAC hardware, and perform testing throughout the design workflow. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Development of functional requirements specification for digital instrumentation and control systems upgrades used at nuclear power plants (NPPs) - 396

Recent market conditions have driven the nuclear industry to become more innovative at reducing cost in order to remain competitive. Instrumentation and Control (I and C) systems used by many NPPs are based on outdated 1960's and 70's technology. Many of the components used in these digital systems are no longer supported by Original Equipment Manufacturers (OEMs) and are rapidly becoming obsolete. Coping with the high cost of replacing I and C equipment due to obsolescence adds to this economic challenge. While other industries have migrated from analog to modern digital I and C technology, implementation of digital I and C controls within the nuclear industry has been very slow. This is primarily due to the high capital cost related to implementation of digital I and C upgrades, which is compounded by regulatory uncertainties that present significant licensing risks. The Nuclear Regulatory Commission (NRC), Nuclear Energy Institute (NEI) and the industry's Digital I and C Working Group are working to address many of the regulatory barriers that adversely impacts implementation of digital upgrades at NPPs. The NRC has developed an 'Integrated Action Plan to Modernize Digital Instrumentation and Controls Regulatory Infrastructure' (ML16097A182), Reference 1. Once fully implemented, this plan is expected to help reduce regulatory uncertainties that affects project cost. Excluding regulatory uncertainty, most OEM's will concur that one of the most important factors in reducing project cost is assuring that all requirements are well defined and complete prior to starting the project. A detailed and high quality Functional Requirements Specification (FRS) is critical in documenting these requirements. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

New intrinsically smart severe accident instrumentation saves costs and enhances severe accident management - 302

The Severe Accidents (SA) at Fukushima in 2011, identified the need for enhanced I and C to follow the progress of SA. Since SA environments are significantly Beyond DBA conditions, some new advanced Inherently Smart sensors had to be developed and tested. Nuclear plant designers and operators have helped focus the I and C enhancement based upon lessons learned from the Fukushima accidents. The result has been the addition of I and C for direct measurement of SA conditions hydrogen concentration, oxygen concentration, temperature, and pressure that provide unambiguous information for management of Severe Accident scenarios. The benefits are immediate from three important aspects. First, they deliver on the nuclear promise for enhanced safety at reduced costs. These Inherently Smart sensors have no maintenance and replace equipment that has high maintenance costs. Secondly, these sensors provide multiple benefits, such as condition monitoring information all the time and direct measurement of SA conditions if they are needed. Thirdly, and most importantly, simplify operator actions and training for SA. SA scenarios and SA Management can now rely on direct measurements to know such things as, the extent of fuel damage, integrity of the reactor pressure vessel, when corium is on the floor of containment, and conditions upon which to activate the hardened vent. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Safety software release options to support plant start-up and commissioning activities - 285

The schedule demands on new nuclear power plant site testing and commissioning programs challenge the plant's Instrumentation and Control (I and C) supplier's ability to provide certified safety software releases to meet the needs of these programs. This paper outlines the typical certified safety software development and certification process which follows a Waterfall software life-cycle model. The conflicts that result between supporting the site program and the safety software supplier's internal certification program are identified. An adapted 'functional' safety release process is presented that supports the site program's needs for software changes and in parallel enables the safety software certification process to proceed. Comparisons between the two safety software release processes are examined and the benefits realized through the use of the adapted 'functional' safety software release process are discussed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Impact of Passive Safety on FHR Instrumentation Systems Design and Classification

Fluoride salt-cooled high-temperature reactors (FHRs) will rely more extensively on passive safety than earlier reactor classes. 10CFR50 Appendix A, General Design Criteria for Nuclear Power Plants, establishes minimum design requirements to provide reasonable assurance of adequate safety. 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, provides guidance on how the safety significance of systems, structures, and components (SSCs) should be reflected in their regulatory treatment. The Nuclear Energy Institute (NEI) has provided 10 CFR 50.69 SSC Categorization Guideline (NEI-00-04) that factors in probabilistic risk assessment (PRA) model insights, as well as deterministic insights, through an integrated decision-making panel. Employing the PRA to inform deterministic requirements enables an appropriately balanced, technically sound categorization to be established. No FHR currently has an adequate PRA or set of design basis accidents to enable establishing the safety classification of its SSCs. While all SSCs used to comply with the general design criteria (GDCs) will be safety related, the intent is to limit the instrumentation risk significance through effective design and reliance on inherent passive safety characteristics. For example, FHRs have no safety-significant temperature threshold phenomena, thus enabling the primary and reserve reactivity control systems required by GDC 26 to be passively, thermally triggered at temperatures well below those for which core or primary coolant boundary damage would occur. Moreover, the passive thermal triggering of the primary and reserve shutdown systems may relegate the control rod drive motors to the control system, substantially decreasing the amount of safety-significant wiring needed. Similarly, FHR decay heat removal systems are intended to be running continuously to minimize the amount of safety-significant instrumentation needed to initiate operation of systems and components important to safety as required in GDC 20. This paper provides an overview of the design process employed to develop a pre-conceptual FHR instrumentation architecture intended to lower plant capital and operational costs by minimizing reliance on expensive, safety related, safety-significant instrumentation through the use of inherent passive features of FHRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Human Performance Measures for the Evaluation of Nuclear Power Plant Control Room Interfaces: A Systematic Review - 265

The purpose of this study was to review the literature related to measuring human performance in the design and evaluation of the human-system interface (HSI) of nuclear power plant (NPP) control rooms and synthesize lessons learned and future directions for research and practice in this field. The design and evaluation of digital or hybrid HSI requires comprehensive measurement of human performance to ensure quality and safety of the nuclear power process control. Human factors evaluation frameworks have been proposed to guide the measurement of human performance constructs. However, there are few guides for selecting appropriate measures for those constructs. A systematic review of empirical research literature was conducted. Studies were identified through academic publication databases and grey literature searches. Eligible studies were coded and information regarding human performance measures was extracted and analyzed. A total of 80 studies met the criteria. Human performance measures extracted from the reviewed studies included measures of: behavioral performance; workload; situation awareness; human-human or human-machine cooperation; other cognitive performance indicators; and system performance. Behavioral performance (speed and accuracy), workload, and situation awareness were the most frequently measured constructs with validated and domain-specific instruments. The results of the literature review provided an overview of the human performance constructs and their corresponding measures for NPP control room HSI evaluation. These results can be used as an initial guide to establish appropriate and comprehensive measurements of human performance for such evaluations. Future directions in measure development are also discussed. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗