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 307 records · Page 17
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Development and implementation of an in-situ cable condition monitoring method for nuclear power plants - 130

The sustainability of the existing fleet of nuclear power reactors depends on utilities' ability to manage the aging of key components that cannot easily or economically be replaced or upgraded. Cables, metals, and concrete structures are among the major and key components that the industry and government are working on to ensure the long term reliability and safety of nuclear plants. To ensure that these key components continue to meet their performance requirements, new testing and condition monitoring technologies must be developed that are capable of identifying and quantifying degradation in these materials. The work presented in this paper focuses on the development and implementation of an in-situ condition monitoring method for nuclear power plant cables. This cable condition monitoring method is based on the frequency domain reflectometry (FDR) technique, which is a non-destructive in-situ electrical test that uses the principle of transmission line theory to locate and quantify impedance changes in a cable circuit. Through extensive research, development, and validation, the FDR test technique has been adapted as a cable condition monitoring method that trends with increasing cable degradation. In addition, the FDR data has been correlated to the industry standard elongation at break (EAB) test to quantify the degree of localized degradation in nuclear plant cable insulation material caused by long term exposure to high temperatures and radiation. By correlating FDR to EAB, the aging condition of cables routed through harsh environments can be better understood in relation to its ability to perform its intended function. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Radiation hardened RF transceiver for in-containment environment applications using commercial off the shelf components - 154

The ability to use wireless devices and other electronics in a radiation environment such as Nuclear Power Plant containment is mostly limited by the radiation sensitivity of these electronics. The benefits derived from wireless technology such as the reduction of the amount of cable needed for new sensors, the reduction of the labor associated with cable routing and the reduction or elimination of containment penetrations for new sensors is significant and well understood; unfortunately these benefits have not been fully accomplished yet because of numerous challenges, one being the radiation environment. Functional operation and radiation effects of a transceiver, which utilizes an integrated circuit system-on-chip, with a micro controller, serial driver, phase-locked loop, crystal oscillators and an RF amplifier, will be described. The second type of transceiver that will be discussed uses discrete components such as JFETs, MOSFETs and op-amps to perform the signal conditioning and RF transmission is also described. Operation capabilities and radiation effects for these two approaches as well as energy sources will be described and compared in detail. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Fault diagnosis and severity estimation in nuclear power plants - 196

Signed Directed Graph (SDG) is a type of fault diagnosis method with a capability to display the complex relationship between plant parameters and reveal the potential danger and the propagation path when a fault occurs in a nuclear plant. However, the fault diagnosis method based on qualitative SDG model inevitably has a number of disadvantages, and some of these disadvantages include its inability to identify reoccurring faults; ambiguous reasoning that can result in low resolution; performance that relies on the accuracy of the model; slow diagnosis rate for large systems and poor real-time diagnosis, among others. In this paper, we propose an SDG fault diagnosis methodology based on Granular Computing Theory, referred to as Granular Computing Signed Directed Graph (GCSDG). This paper shows how knowledge reduction nature of granular computing based fault diagnosis can simplify the decision table, reduce the allocation of resources, scale-down the complexity of the problem and improve the efficiency of data processing. Then, we estimate the severity of the fault using Back Propagation (BP) neural network approach. The proposed methodology was verified by considering certain faults in the primary loop of a PWR, using PCTRAN. The result shows a considerable improvement in fault diagnosis and the severity estimated serves as a useful information for the operator, providing a firm foundation for further intervention. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Improving plant efficiency through the installation of TecSOLCEP online monitoring system - 362

In February 2016, Nuclear Energy Institute (NEI) launched its 'Delivering the Nuclear Promise (DNP)' strategic plan with the objective of reducing electric generating costs by 30% across the nuclear power industry by 2018. Improving efficiency has been addressed as one of the strategic focus areas in the plan. In nuclear power plants (NPPs) there is room for improvement in main components efficiency and TecSOLCEP is a software tool that has proved its value for that purpose. TecSOLCEP is a software tool for monitoring, diagnosing and quantifying performance deviations in Power Plants. It faithfully follows the methodology of ASME standards. Over 15 years of success stories in more than 50 electric power plants of different technologies worldwide, justify the benefits for the plants. The implicit concept in TecSOLCEP is 'Benefit based Maintenance' (BbM) as complementary to usual maintenance strategies like time-based, condition-based or predictive. These maintenance strategies are oriented to obtain the Safety and Reliability goals shared by all the electricity generation technologies. BbM is oriented to get the maximum economic benefit by Efficiency care, which historically has not been a priority, especially in nuclear plants. BbM applied with the help of performance monitoring tools like TecSOLCEP, is called to play an important role to improve efficiency which, in plants mainly operating at full power, is equivalent to increase the maximum power. Nuclear power plants that intend to extend their lives may also take advantage of identifying and quantifying inefficiencies in equipment to better address necessary investments. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Overview of simulation and control tools for extended operability of nuclear reactors - 183

The increasing penetration of Renewable Energy Sources (RES) requires the current power grid to be suitably improved. Because of the unsteady nature of RES, as long as energy storage facilities are not deployed and commercialized, an increased effort for the frequency regulation in order to ensure high reliability and performance is necessary. In such a scenario, nuclear units (traditionally deployed as baseload generating units) will be required to cooperate with intermittent RES-based power plants. Therefore, new reactors will need larger operational flexibility with respect to the capabilities of the currently operated units. Over the last few years, a comprehensive research program oriented at developing a suite of simulation tools and control system development capabilities was pursued at Argonne National Laboratory. In addition to the study of advanced reactor concepts (e.g. metal-fueled Sodium-cooled Fast Reactors) and technological innovations (such as S-CO{sub 2} Energy Conversion Cycles), new control approaches were developed to improve the operability of nuclear power plants. In particular, Model-based Predictive Control algorithm for Multiple Input Multiple Output control strategies and set-point definition algorithm based on Reference Governor were studied and applied to specific scenarios, providing with a range of novel capabilities able to meet the expected greater sets of requirements mentioned above. In this work, an overview of such tools and case studies is presented with particular focus on their impact on operability and economic competitiveness of nuclear energy. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Framework for structural online health monitoring of aging and degradation of secondary piping systems due to some aspects of erosion - 330

This paper describes the current state of research related to critical aspects of erosion and selected aspects of degradation of secondary piping systems in nuclear power plants (NPPs). The paper also proposes a framework for online health monitoring of aging and degradation of secondary piping. The framework consists of an integrated multi-sensor modality system, which can be used to monitor different piping configurations under different degradation conditions. The paper analyses the currently known degradation mechanisms and available monitoring techniques and practices. Based on this analysis, the structural health monitoring framework is proposed. There is unanimous agreement between industry experts and academic researchers that identifying and prioritizing inspection locations in secondary piping systems would eliminate many excessive in-service inspections. The proposed structural health monitoring framework takes aim at answering this challenge by combining long range guided wave technologies with other monitoring techniques, which can significantly increase the inspection length and pinpoint the locations that degraded the most. More widely, the paper suggests research efforts aimed at developing, validating, and deploying online corrosion monitoring techniques for complex geometries, which are pervasive in NPPs This transition to condition-based, risk-informed automated maintenance will contribute to a significant reduction of operations and maintenance costs that account for 80% of nuclear power generation costs. To address the issue of rising operating costs and economic viability, in 2017, companies that operate the national nuclear energy fleet started the Delivering the Nuclear Promise Initiative, which is a 3-year program aimed at maintaining operational focus and safety, increasing value, and improving efficiency. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Time-of-flight Extreme Environment Diffractometer at the Helmholtz-Zentrum Berlin

The Extreme Environment Diffractometer (EXED) is a new neutron time-of-flight instrument at the BER II research reactor at the Helmholtz-Zentrum Berlin, Germany. Although EXED is a special-purpose instrument, its early construction made it available for users as a general-purpose diffractometer. In this respect, EXED became one of the rare examples, where the performance of a time-of-flight diffractometer at a continuous source can be characterized. In this paper, we report on the design and performance of EXED with an emphasis on the unique instrument capabilities. The latter comprise variable wavelength resolution and wavelength band, control of the incoming beam divergence, the possibility to change the angular positions of detectors and their distance to the sample, and use of event recording and offline histogramming. These features combined make EXED easily tunable to the requirements of a particular problem, from conventional diffraction to small angle neutron scattering. The instrument performance is demonstrated by several reference measurements and user experiments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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 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 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↗