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 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Radiation Hardened Electronics for Reactor Environments

Radiation-hardened (rad-hard) electronics have been the shortcoming in nuclear sensing and instrumentation. Placing advanced sensors and associated electronics closer to a nuclear reactor core will improve reactor control and operation through increased signal accuracy, precision, and fidelity, resulting in safer, more efficient energy production. However, the extreme temperature and bombarding radiation environment of nuclear applications make research and development into electronics and electronic materials technologies essential to enabling improved safety, monitoring, and control of the existing nuclear reactor fleet and the next generation of reactors, including microreactors. To help the US Department of Energy (DOE) define a course for the Office of Nuclear Energy (NE)-funded rad-hard electronics research, Oak Ridge National Laboratory staff present this summary of the state-of-the-art (SOA) technology for the design and implementation of rad-hard electronics and systems, including research efforts and commercial offerings. A collection of technological gaps associated with reactor applications are compiled herein. Based on the radiation degradation mechanisms discussed in this report, along with the state-of-the-art summary survey and a compiled list of gaps, a high-level research plan of future NE funding direction was developed and is provided. This report outlines these activities and provides a summary of SOA electronics technology for rad-hard instrumentation in reactors, a list of the major technology gaps observed, and a draft plan for addressing the present and future needs of reactor instrumentation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Considerations of Radiation-Hardened Electronics for Alpha Detectors in Molten Salt Reactors

Along with advanced generations of reactors quickly becoming reality, the need for instrumentation to provide reliable information for their safety and operation is also becoming relevant. Instrumentation-wise, molten salt reactors (MSRs) represent an especially challenging case in which detectors must survive extreme temperatures, high amounts of radiation, and a highly corrosive coolant. Adding to the complexity, tight material control and accounting must be maintained with any special nuclear material (SNM), a practice made more difficult by the nature of a loop of flowing fuel. Alpha spectroscopy offers a unique window into investigating the isotopic concentrations of actinides within just such a molten salt environment. The characteristic α emissions of specific actinide isotopes in the salt enable quantification of their concentrations. Semiconductor devices based on wide bandgap materials such as silicon carbide and gallium oxide can withstand the extreme temperatures within MSRs and offer good chemical stability with the corrosive salt. Collaborators at The Ohio State University are fabricating and testing a β gallium oxide α particle sensor that could identify actinide concentrations within the fuel salt to help improve the material control and accounting required for complicated system inherent to MSRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

Guidance Regarding Probable Approaches and Instrumentation to Liquid Fuel Molten Salt Reactor Material Control and Accounting (MC&A) Rev. 1

With the rapid development of advanced reactors and numerous companies requesting pre-application engagements with the Nuclear Regulatory Commission (NRC), there is need for the NRC to best prepare themselves for new challenges presented by advanced technologies. One specific reactor type of interest is the molten salt reactor (MSR) that uses liquid salt as the fuel for the reactor. Liquid salt fuel reactors have specific challenges with regards to performing material control and accounting (MC&A). The challenges result from the fuel form being a continuous fissile material form rather than discrete units like is found with fuel assemblies in traditional light water reactors (LWRs). The goal of this document is to provide the NRC with guidance on the different approaches and technologies that may be proposed by reactor designers to address these problems to prepare staff for regulatory reviews.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Common cause failure (CCF): a path to quantitative success - 134

In the development of I and C systems for nuclear plants it is critical to validate a defense in depth strategy through a thorough and complete reliability analysis program. An emerging issue is how to define and use a quantitative estimation of CCF in a system analysis. Industry standards provide insight but not necessarily a comprehensive approach. One solution is to leverage existing modeling techniques with a unique insertion of CCF probabilities. A relatively simple approach will be provided using reliability block diagrams. Three specific areas that will be explored include: 1. Definition of hardware CCF; 2. Quantitative estimation in a simple system; 3. Quantitative estimation in a complex system. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Pruning of Discrete Dynamic Event Trees Using Density Peaks and Dynamic Time Warping

Discrete Dynamic Event Trees (DDETs) are used in probabilistic risk assessment (PRA) to determine the risk associated with complex systems, such as nuclear power plants, where accident scenarios need to be realistically modeled. One of the most challenging aspects of the DDET method is that a large number of branches are produced for each initiating event. This can result in data that is difficult to organize and time consuming to analyze. As the number of uncertain parameters increases, so does the complexity of the event tree. An event tree often has nodes that do not contribute to the overall results and predictive capabilities; therefore, we suggest a pre-pruning algorithm that can work to remove these nodes, and consequently, reduce the complexity of the tree. Sandia National Laboratories (SNL) is currently working on a two year dynamic cyber-risk informed project to assess cyber failure in a nuclear power plant. This will be achieved by coupling SNL's dynamic scheduler known as Analysis of Dynamic Accident Progression Trees (ADAPT) with MELCOR, a reactor accident simulator, and a System for Circuit Evaluation and Prediction of Transient Radiation Effects (SCEPTRE), a cyber Emulytics code. This paper will discuss the preliminary results of a feasibility study for using density peaks (DP) clustering with dynamic time warping (DTW) to prune branches that are formed from large DDETs, like the ones that will be produced throughout SNL's project. Papers from the project that are submitted to this conference discuss the following: - The MELCOR model for a Pressurized Water Reactor that uses a digital instrumentation and control system. - Proposed generic cyber exploits for the above reactor. - Modifications to ADAPT to extend the framework for multiple simulators, incorporate importance measures into DDET analysis, as well as sort and reduce DDETs based on inputs to or outputs from the tree. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Dynamic attack surfaces in nuclear power plants - 162

In this paper, we will summarize a group of architectural principles that inform the development of secure control system architectures, followed by a methodology that allows designers to understand the attack surface of components and subsystems in a way that supports the integration of these surfaces into a single attack surface. We will then show how this methodology can be used to analyze the control system attack surface from a variety of threats, including knowledgeable insiders. We close the paper with an overview of how this approach can be folded into a more rigorous mathematical analysis of the system to define the system's security posture. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Virtualized hardware environments for supporting digital I and C verification - 324

The technology of virtualized hardware testing platforms within the nuclear context in order to provide a framework for evaluating virtualized hardware/software approaches is discussed. To understand the applicability of virtualized hardware approaches, we have developed a virtual hardware model of an actual smart sensor using the Imperas product OVPsim development environment, as is seen in . An evaluation of the modeling effort required to create the digital I and C smart sensor is provided, followed by an analysis and findings on the various tools in the OVPsim environment for supporting testing, evidence collecting, and analysis. The analyzed tools include traditional input/output testing, model-based testing, mutation testing, and fault injection techniques. Finally, the goal of this research effort is to analyze virtual hardware platforms and present findings to the nuclear community on the efficacy and viability of this technology as a complementary means to actual hardware-in-the-loop testing Factory Acceptance Testing for Nuclear Power Plant (NPP) digital I and C systems. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

A new international standard on cybersecurity for nuclear power plants: IEC 62645 - cybersecurity requirements - 161

This paper provides an overview of the IEC 62645, an actual standard by the International Electrotechnical Commission (IEC) focused on the issue of requirements for computer security programmes and system development processes to prevent and/or minimize the impact of cyber attacks against computer-based I and C systems. Developed since 2009 and published end of 2014, the standard is intended to be used for changing or establishing new security programmes for I and C systems of operating and new Nuclear Power Plants (NPP). This paper also presents the key issues being considered in the new revision to IEC 62645, currently in process. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Condition monitoring and false alarm reducing of sensors in NPP - 222

With more and more application of digital I and C systems in nuclear power plants (NPP), more and more sensors are also required. Large amount of sensors really can make contribution to the condition monitoring of systems and equipment in NPP, whereas the heavy use also increases the malfunction risk of sensor self. Therefore the reliability and effectiveness of sensors must be guaranteed first. Thus the condition of sensors should be monitored first in a NPP. Since the precise mathematical models of sensors in complex systems are relative difficult to establish, however quantity of operating data can be easily obtained due to the wide application of digital I and C systems, thus this article adopts principal component analysis (PCA) to carry out condition monitoring for sensors. The PCA model developed for sensors is proved to be effective with operating data acquired from a real NPP. It can detect and isolate the fault when an artificial fault is imposed to the normal data. Then the monitoring results are further processed by another control limit (second confidence limit) besides T 2 and Q statistics (first confidence limit) in PCA. The value of second confidence limit is calculated based on the probability statistics of T 2 and Q statistics. And simulation tests show that the application of the second confidence limit can greatly reduce false alarms in the PCA model and the model performance can be greatly improved with the application of false alarms reducing method. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Manual reactor power oscillations were observed in real-time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Complexity and error potential of digital I and C - 135

Utilization of digital I and C for application in nuclear power plants led to a growing discussion on how to create methods for a probabilistic evaluation on such systems. Issues arise on the one hand due to a lack of commonly accepted models for quantitative evaluation of availability and reliability and on the other hand due to the integration of such models into the evaluation of reliability of an overall system. Nevertheless programmable and even more complex systems are implemented to satisfy growing functionality requirements. Therefore the question for a method to evaluate the reliability of systems based on software or on programmable logic became even more urgent. The approach described is based on the commonly accepted assumption that more complex designs are less reliable. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Qualifying the Instrument Thimble 11 Test Position in the Advanced Test Reactor Critical

The Advanced Test Reactor (ATR) is a versatile nuclear research reactor located at the Idaho National Laboratory (INL) in Eastern Idaho, United States of America. It is one of the most powerful and flexible research reactors in the world and is primarily used for materials testing, isotope production and basic nuclear science research. The ATR is a light-water-cooled, beryllium-moderated reactor with a nominal thermal power capacity of 250 MW utilizing specially designed fuel, arranged in a serpentine pattern to create 9 flux traps. The special design of ATR allows for the neutron flux within each of the flux traps providing flexibility for experiments, programs and allowing experiments with different dose requirements to be irradiated simultaneously. The Advance Test Reactor Critical (ATRC) is a full-scale replica of the ATR core but is housed in a pool instead of a pressure vessel. ATRC is usually operated at less than 600 watts [1]. There are 12 dry instrument thimbles located around the outside of the core tank where a variety of measurement equipment is housed to facilitate reactor operations. Currently Instrument Thimble 11 (IT-11) is not utilized in ATR or ATRC. A range of measurements is ongoing to qualify IT-11 as a test platform for nuclear instrumentation research, development and further experimentation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Identification of the accident-related critical digital assets based on probabilistic safety analysis results - 225

Digital I and C systems have been developed and installed in nuclear power plants, and due to installation of the digital I and C systems, cyber security concerns are increasing in nuclear industry. However, there are too many critical digital assets (CDAs) to be inspected in digitalized NPPs. In order to reduce the inefficiency of regulation in nuclear facilities, the critical digital assets that are directly related to an accident are elicited by using the probabilistic safety assessment results. Target initial events are selected, and their heading are analyzed through event tree analysis. Among the headings, the headings that can be proceeded directly to the core damage by the cyber-attack are finally selected as the target of deriving the minimum cut-sets. Based on success criteria of each heading, we analyze the fault trees and derive the minimum set-cuts. In terms of original PSA, the value of probability for the cut-sets is important but the probability is not important in terms of cyber security of NPPs. The important factors is the number of basic events consisting of the minimal cut-sets that is proportional to vulnerability. Finally, the process for identifying accident-related CDAs is suggested based on PSA results. The results of this study are expected to be used to derive the linkage between cyber-attack and accident, and to develop the final core digital asset identification methodology and effective regulatory method. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Workload Response to Soft Controls Presented on Two Interfaces - 382

Most operational Nuclear Power Plants (NPP) in the United States are primarily outfitted with physical (analog) Instruments and Controls (I and C). However, as parts wear out in legacy plants, these I and Cs are being replaced by digital interfaces. Furthermore, new plants are being proposed and built domestically and throughout the world that are composed primarily of soft controls in the main control room. To that end, it is important to understand if and how various types of interfaces affect performance. The present effort sought to compare a desktop and touchscreen interface. The Instantaneous Self-Assessment, the NASA-Task Load Index, and electrocardiography were used to assess workload response while performing three main control room task types on each interface. Results have implications for interface selection and assessment of new NPP main control room designs. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1973

Nuclear Science Symposium, 19th, and Nuclear Power Systems Symposium, 4th, Miami, Fla., December 6-8, 1972, Proceedings.

Major topics covered include radiation monitoring instrumentation, nuclear circuits and systems, biomedical applications of nuclear radiation in diagnosis and therapy, plasma research for fusion power, reactor control and instrumentation, nuclear power standards, and applications of digital computers in nuclear power plants. Systems and devices for space applications are described, including the Apollo alpha spectrometer, a position sensitive detection system for UV and X-ray photons, a 4500-volt electron multiplier bias supply for satellite use, spark chamber systems, proportional counters, and other devices. Individual items are announced in this issue.

Source record↗
DOE OSTI2017

NBSR reactor control room upgrade - 215

The maintenance and modernization of the National Institute of Standards and Technology (NIST) Center for Neutron Research (NCNR) reactor, known as the NBSR, is an ongoing process to endure until the eventual decommissioning of the reactor. As part of the renovation effort, the reactor control system is to be upgraded. Following some initial outsourcing efforts, the NBSR reactor control system upgrade has been restructured to be implemented by in-house means. The beginning stage of the upgrade involved the development of a simulator environment based on a PC system and digitization of the majority of nuclear reactor monitoring data. The NBSR reactor surveillance data consists of approximately thirteen hundred input/output points, including analog/digital sensor signals, signal conditioners, and relay logic. The aim of the current phase is to implement a new reactor data acquisition and display system (RDS) and lay out the infrastructure for the upgrade of the reactor control system (RCS). Key design criteria are to implement a reliable, redundant, diverse, modular and candid system where any complex digital manipulations are taken outside of the control room. Basically, the data analytics and custom codes are separated from reactor operation control and decision making logic. Furthermore, the new RDS is designed such that it could be maintained internally by NCNR personnel. The RDS is unique in design and modularity, minimizing custom codes and testing, making it ideal for research reactors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Model-based verification of I and C specifications - 184

In the field of nuclear I and C (Instrumentation and Control), the late detection of errors in the specifications and the design of I and C has a strong impact in terms of time and money. The early verification in the V cycle of I and C functions is a key issue to capture inconsistencies and errors before the implementation and on-site tests. EDF, in the framework of the collaborative project CONNEXION, has investigated methods based on simulation to formalize I and C requirements and to evaluate them at the conceptual and basic design phases. Behavioral models of the physic parts of the system and I and C functions, described as functional diagrams, are co-simulated in an integration environment based on the FMI (Functional Mockup Interface) standard. In this specific environment, test scenarios are performed on these models. These scenarios are defined on the basis of a high level model of the studied system that takes into account the states to cover (in normal and degraded situations). Requirements, once formalized, are translated into observers that evaluate their fulfillment during the simulations. Simulation logs are then a tool for the designer to analyze the causes of the possible non-fulfillment of requirements and to fix the corresponding errors in the specifications. The proposed methodology enables to detect errors in the specifications of I and C functions, to generate automatically functional test scenarios taking into account the coverage of the requirements and all the different operational modes of the system. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗