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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.

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At least 163 records · Page 9
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Identification and repair of intermittent cable faults in nuclear power plants - 131

The nuclear industry uses cable circuits as an integral part of a plant's reactor protection system (RPS), instrumentation and control (I and C) and power distribution systems. The reliability of these systems is critical for continued operation of the world's aging nuclear fleet, as well as for safe shutdown of the reactor. When intermittent faults occur in cable circuits, the symptom of the problem often appears and disappears in milliseconds while the aftereffects can result in spurious alarms and automatic actuation of safety systems. Problems are compounded when component degradation causes a cable circuit to transmit inaccurate information that does not represent actual plant operating conditions. The root cause of intermittent cable circuit operation can include degradation of cable and connectors from heat, humidity, radiation, and vibration. Unlike hard faults (which are constant open or short circuits), intermittent faults are defined as a temporary change in impedance that may only last for a few milliseconds. While less disruptive than hard faults, signal spiking from intermittent faults can often lead to alarm conditions and reactor trips. A soft fault is defined as a subtle variation the electrical characteristics of the circuit that still permit the continued operation of the circuit. A soft fault may be a precursor to a hard fault due to continued degradation of the fault over time. The focus of this paper is the description of a cable testing system that can be connected to an active, low frequency circuit and continuously monitor electrical characteristics using the time domain reflectometry (TDR) technique until an intermittent impedance change occurs. This technique uses the difference in frequency response of the circuit and the TDR to isolate the TDR signature from the active circuit data . When a momentary impedance change occurs that exceeds a preconfigured limit threshold, the distance to fault data is automatically overlaid with the healthy baseline reference and stored in a database for diagnosis and repair. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Digitalization of an experimental electrochemical reactor via the smart manufacturing innovation platform

The exponential increase in data produced over the last two decades has revolutionized the way we collect, store, process, analyze, model, and interpret information to improve profitability. Manufacturing is no exception. How- ever, Smart Manufacturing, the digital practice, organization, workforce, and infrastructure transformation for collection and deployment of data and models at scale and at all levels of manufacturing, is a complex, costly, and labor-intensive journey that is still seeing slow adoption. The Clean Energy Smart Manufacturing Innovation Institute (CESMII), a national Manufacturing USA public-private partnership sponsored by the Department of Energy, is addressing this scaled use of data and modeling in manufacturing. CESMII has focused on how to col- lect and use operating data for numerous applications that improve productivity, precision, and performance of manufacturing operations from factory floor to supply chain using process simulation, predictive analytics, mon- itoring and control, and real-time optimization. Because contextualized data are key, CESMII has developed the Smart Manufacturing Innovation Platform (SMIP) to lower the barriers to the data that are needed to accelerate data-based model building, improve data visualization, and more quickly gain insights. Reusable, standards-based ways of doing data collection, ingestion, and contextualization are particularly important for scaling access and use of data. The SMIP uses a standards-based definition and construct for reusable information models called an SM Profile. When an SM Profile is used in conjunction with the SMIP, the SMIP ensures the availability of contextualized, operational data for model building. The present work demonstrates Smart Manufacturing and the application of the SMIP for building several data-centered models for the operation and control of an ex- perimental electrochemical reactor that reduces carbon dioxide (CO 2 ) gas to valuable liquid and gas chemicals, such as alcohols, olefins, and syngas. We describe how the SMIP plays a central role in more effective model building and we demonstrate how the electochemical reactor can be controlled and optimized for the desired products. Use of the SMIP involves the transmission of real-time sensor measurements to a cloud resource so that the operating data are available to all model building experts. The data collection and transmission process is fully automated to greatly reduce the need for manual manipulation of the data. Data-driven machine learning models are used for advanced real-time state estimation, real-time optimization, and model-based feedback control for the reactor. The application models are implemented as a system to monitor the data flow and control the electrochemical reactor with a single visualization interface. SM Profiles are used to demonstrate reusability of the information models for the reactor and the instrumentation. The application packages, algorithms, and user interfaces developed are cast as Docker images in a library to facilitate reusability of the application models.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Commercial Design of Custom Front-end Electronics for a High Temperature Fission Chamber

The next generation of advanced nuclear reactor will require instrumentation designed to withstand higher temperatures and harsher environments. One of the most important instruments required to monitor power and in-core fuel management is a fission chamber. A fission chamber consists of a gas-filled chamber with a fissile coating on the inside of the chamber.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Testing of Instrumentation and Control Sensors and Cables for Small Modular Reactors

As advanced reactor concepts like small modular reactors (SMRs) progress towards design maturity and commercial deployment, it is important to assess their instrumentation and control (I&C) system sensors and cables to assure their safe, reliable, and efficient operation throughout their service lifetimes. The I&C system of any nuclear power plant (NPP) is the central nervous system of the plant and is made up of field devices such as sensors which are connected via cables to analog and/or digital systems. These systems are responsible for control and/or protection of the plant. The performance of I&C sensors and cables is dependent on several factors, including plant operating conditions. In particular, the plant operation can affect the static (accuracy) and dynamic (response) performance of the sensor, whereas the surrounding ambient environment can impact the remaining useful life of the cables, connectors, and other associated components within I&C sensor circuits.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Complete Evaluation on Advanced Reactor Machine Learning Subversion Attacks (Final)

Navigating through the world of Artificial Intelligence (AI) in nuclear reactors and their Instrumentation and Control (I&C) systems demands a careful, deliberate journey. AI’s capability to manage massive datasets and streamline control systems has indeed carved out a significant role in various sectors, including nuclear energy. However, while AI, and particularly Large Language Models (LLMs), bring a lot to the table in terms of operational efficiency and anomaly detection, they also expose the sector to a new breed of cybersecurity threats, like Inference Attacks, Adversarial Attacks, and Trojan Attacks. This guide is designed to be a straightforward manual, diving deep into the intertwining worlds of AI and cybersecurity within nuclear reactors, and tailoring insights for three crucial audiences: I&C Vendors/Developers, Nuclear Regulators, and Nuclear Reactor Operators and Cyber Defense Teams. (1) Section 2, directed at I&C Vendors/Developers, will provide a clear and focused look at several cybersecurity attacks, offering practical recommendations and detailed scenarios related to AI cybersecurity. This section isn’t just about identifying problems but also about giving solid, usable solutions. (2) Section 3, meant for Nuclear Regulators, gets straight to the point about regulations, policy suggestions, and guidelines that are needed to lay down a robust, secure, and ethical foundation for the application of AI in nuclear operations. The focus is on making sure that everything adheres to international standards and laws while being practicable and clear-cut. (3) Section 4, aimed at Nuclear Reactor Operators and Cyber Defense Teams, offers an exhaustive exploration and technical reports, with clear recommendations and scenario analyses vital to protect operational environments and guarantee the secure application of AI in nuclear reactor operations. The goal is simple: as we step into an era where AI becomes a fundamental element of our technological and energy infrastructures, this guide is here to act as a clear, direct handbook, ensuring that AI is implemented within the nuclear sector in a manner that is secure, responsible, and practical. It’s about striking a balance – optimizing the undeniable benefits offered by AI while securing and shielding against potential cyber threats as we move through this new and complex landscape.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Bayesian inference for high confidence signal validation and sensor calibration assessment - 328

Online Monitoring (OLM) systems has been investigated for applications in the nuclear industry for the past two decades. These systems offer economic and technological advantages over the currently implemented time based testing methods for sensors and associated instrumentations. The advantages offered not only augment economic viability of current nuclear reactors but will be an inevitable change called for to suit the technologies of the advanced reactors (Gen III+). Irrespective, the acceptance of OLM into the regulatory structure in the U.S. is still hindered by technical and structural clarity on uncertainty quantification used by these systems. A Bayesian inference technique is implemented in the form of a residual model to quantify uncertainty in OLM predictions. The developed system used Gaussian processes to represent the OLM predictions and model inadequacy, the stationary representation of model based uncertainty. In addition, model selection techniques are used in conjunction with numerical methods such as Genetic Algorithm (GA) and Gibb's sampler to add robustness to the estimation model inadequacy, particularly in the presence of drifts and faulted conditions. The prediction model developed herein find applicability in fault detection, obtaining real values of processes after the onset of faults etc. Further development looks into the extension of current method for a multiple output scale implementation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1982

Low and medium heating value coal gas catalytic combustor characterization

Catalytic combustion with both low and medium heating value coal gases obtained from an operating gasifier was demonstrated. A practical operating range for efficient operation was determined, and also to identify potential problem areas were identified for consideration during stationary gas turbine engine design. The test rig consists of fuel injectors, a fuel-air premixing section, a catalytic reactor with thermocouple instrumentation and a single point, water cooled sample probe. The test rig included inlet and outlet transition pieces and was designed for installation into an existing test loop.

Schwab, J. A.↗
NASA NTRS1991

Report on the formal specification and partial verification of the VIPER microprocessor

The formal specification and partial verification of the VIPER microprocessor is reviewed. The VIPER microprocessor was designed by RSRE, Malvern, England, for safety critical computing applications (e.g., aircraft, reactor control, medical instruments, armaments). The VIPER was carefully specified and partially verified in an attempt to provide a microprocessor with completely predictable operating characteristics. The specification of VIPER is divided into several levels of abstraction, from a gate-level description up to an instruction execution model. Although the consistency between certain levels was demonstrated with mechanically-assisted mathematical proof, the formal verification of VIPER was never completed.

Brock, Bishop↗
NASA NTRS1993

Nuclear thermal propulsion technology: Results of an interagency panel in FY 1991

NASA LeRC was selected to lead nuclear propulsion technology development for NASA. Also participating in the project are NASA MSFC and JPL. The U.S. Department of Energy will develop nuclear technology and will conduct nuclear component, subsystem, and system testing at appropriate DOE test facilities. NASA program management is the responsibility of NASA/RP. The project includes both nuclear electric propulsion (NEP) and nuclear thermal propulsion (NTP) technology development. This report summarizes the efforts of an interagency panel that evaluated NTP technology in 1991. Other panels were also at work in 1991 on other aspects of nuclear propulsion, and the six panels worked closely together. The charters for the other panels and some of their results are also discussed. Important collaborative efforts with other panels are highlighted. The interagency (NASA/DOE/DOD) NTP Technology Panel worked in 1991 to evaluate nuclear thermal propulsion concepts on a consistent basis. Additionally, the panel worked to continue technology development project planning for a joint project in nuclear propulsion for the Space Exploration Initiative (SEI). Five meetings of the panel were held in 1991 to continue the planning for technology development of nuclear thermal propulsion systems. The state-of-the-art of the NTP technologies was reviewed in some detail. The major technologies identified were as follows: fuels, coatings, and other reactor technologies; materials; instrumentation, controls, health monitoring and management, and associated technologies; nozzles; and feed system technology, including turbopump assemblies.

Clark, John S.↗
DOE OSTI2021

RELAP5-3D Modeling of High Temperature Test Facility (HTTF)

HTTF at Oregon State University (OSU) Reference: General Atomics? modular high-temperature gas-cooled reactor Helium cooled, electrically heated Prismatic graphite blocks in the core and reflectors Alumina ceramic blocks are used to simulate the core and top and bottom reflectors One-fourth scale in length and diameter Most of the coolant channels in the core are full scale Lower pressure compared to the prototype reactor Over 500 instruments Designed primarily to investigate depressurized conduction cooldown (DCC) transients

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2022

Production of Xenon-135 from isotopically enriched Xenon-134 and Xenon-136 Targets

Atmospheric radioxenon analyses conducted under the Comprehensive Test Ban Treaty Organization (CTBTO) necessitates the collection and analysis of air samples to quantify 135Xe, an indicator for nuclear explosions and reactor operations. Analytical instrument calibration for these analyses requires high quality, monoisotopic reference standards that are typically short lived, and therefore must be produced and shipped promptly to monitoring stations worldwide. This work describes the preparation of 135Xe by neutron capture (n, ?) on monoisotopic 134Xe, neutron irradiation (n,2n) on monoisotopic 136Xe, and high energy photon irradiation (?, n) on monoisotopic 136Xe. The targets are prepared by electromagnetic isotope separation of natural xenon. Results from each process are compared for 135Xe yield and purity.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI1992

Nuclear Safety [Vol. 33, No. 1, January-March 1992]

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: 1 Technical Note: A New Approach to Fission Reactor Safety, Yu. V. Petrov; 5 Erratum; ACCIDENT ANALYSIS: 6 Analysis and Modeling of Fission Product Release from Various Uranium-Aluminum Plate-Type Reactor Fuels, R. P. Taleyarkhan; CONTROL AND INSTRUMENTATION: 23 Applications of a Surveillance and Diagnostics Methodology Using Neutron Noise From a Pressurized-Water Reactor, R. T. Wood, L. F. Miller, and R. B. Perez; DESIGN FEATURES: 36 Westinghouse Advanced Passive 600 Plant, B. A. McIntyre and R. K. Beck; 47 System 80+™ PWR Safety Design, C. W. Bagnal, R. A. Matzie, and R. S. Turk; ENVIRONMENTAL EFFECTS: 58 The MATS Experiments—Mesoscale Atmospheric Transport Studies at the Savannah River Site, A. H. Weber, S. Berman, R. J. Kurzeja, and R. P. Addis; 75 Book Review: Environmental Radioactivity in the European Community 1984-1985-1986, C. A. Little; WASTE AND SPENT FUEL MANAGEMENT: 76 Activities Related to Waste and Spent Fuel Management, M. D. Muhlheim and E. G. Silver; OPERATING EXPERIENCES: 87 Aging Assessment of BWR Control Rod Drive Systems, R. H. Greene; 100 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 103 Selected Safety-Related Events, Compiled by G. A. Murphy; 110 Operating U.S. Power Reactors, Compiled by M. D. Muhlheim and E. G. Silver; RECENT DEVELOPMENTS: 129 General Administrative Activities, Compiled by M. D. Muhlheim and E. G. Silver; 139 Reports, Standards, and Safety Guides, D. S. Queener; 143 Proposed Rule Changes as of Sept. 30 1991; ANNOUNCEMENTS: 128 Harvard School of Public Health Announces Short Courses; 128 16th Biennial ANS Topical Meeting on Reactor Operating Experience: Present and Future Technologies—Applying Lessons Learned (Call for Papers); 147 The Authors; 150 Indexes to Nuclear Safety, Volume 32; 154 Errata.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Modernizing the Legacy Fission Wire Measurement System for the Advanced Test Reactor-Critical Facility

Operational lifetime extensions of existing research reactors have emphasized the need for refurbishment, replacements, and upgrades to supporting equipment and instrumentation. The Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), which entered service in 1967, has recently completed the sixth core internals change-out and has scheduled operations until at least 2040. Reactor maintenance and operational risk management is critically important in the research reactor community, however supporting measurement systems sometimes get overlooked when maintenance is planned. The Fission Wire Measurement System (FWMS) is a custom measurement system designed in the 1960s to measure the beta-particle activity of irradiated uranium-aluminum fission wires. This measurement is conducted to determine the fission rate profile of the Advanced Reactor Test Critical (ATR-C) facility. The ATR-C is an open-pool, low-power test reactor that was purpose driven to resemble ATR and is used to qualify experiment configurations and verify core models prior to full-power experiment irradiations in ATR. A power distribution measurement in ATR-C uses uranium-aluminum wires that are distributed throughout the ATR-C core to validate simulation and modeling results. These measurements require 340 to 1500 wires to be irradiated and measured within a 12-hour window. The activity of the wires is measured in the required time with the FWMS, which was put into service in 1965 at the Radiation Measurements Laboratory (RML). The system consists of 4 measurement channels and one reference channel, each with a 2-pi proportional gas flow detector and the measurement channels each have an automated sample changer. This legacy system is crucial to the continued operations of ATR and has undergone some minor hardware upgrades since 1965, however the system presently relies on custom control boards, custom gas ion chambers, analog amplifiers/discriminators, and a user interface (UI) for the system written in outdated code. Much of the equipment and software is custom with no commercial replacements or support and limited documentation. The existing control software requires an operating system that is no longer supported, creating more vulnerabilities to continued operations. A project is underway with a third-party vendor to design, build, and document a new control and data acquisition system (CDAS) for the FWMS. The new upgrade will replace the control system, computer, UI, sample changer motors, and main power supply while maintaining the interface with existing detector hardware. The upgraded system will be operated in parallel with the current hardware and software to conduct validation testing. This equipment upgrade demonstrates the commitment at ATR to ensuring successful operations and potential future research reactors at INL.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
DOE OSTI2017

Diversity and defense-in-depth analysis for I and C systems of research reactors: A case study on two research reactors - 376

This paper provides the analysis results of the diversity and defense-in-depth (D3) applied to two research reactors designed by the Korea Atomic Energy Research Institute. The D3 assessments for instrumentation and control (I and C) systems of two research reactors are performed according to the guidelines documented in NUREG/CR-6303. The I and C systems are classified into four echelons of defense: control echelon, reactor trip echelon, engineered safety feature (ESF) actuation echelon, and monitoring and indication echelon. The diversity between the echelons of defense is assessed to determine that a single software common cause failure (CCF) does not impair the functions of multiple echelons of defense. In addition, the effects of a software CCF of the reactor protection system on the design basis events of the research reactors are analyzed by assuming that the automated reactor trip and ESF actuation functions of the reactor protection system are inoperable due to the software CCF. From the D3 analyses for the research reactors, it is verified that (1) adequate diversity has been provided, (2) adequate defense-in-depth has been provided, and (3) the displays and manual controls for critical safety functions initiated by operator action are diverse from digital systems used in the automatic portion of the protection system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

MRP Review - MARVEL I&C Overview

MARVEL I&C Overview - The design and status of MARVEL I&C is discussed, along with the safety foundations that constrain and enable the design. Included are outlines of the safety-related I&C components, the non-safety-related I&C components, and the additional elements to be added after dry criticality in preparation for full power operations. Finding the path to remote operations will also be briefly discussed.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗