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 73 records · Page 4

Real-Time In-Pile Acoustic Measurement Infrastructure at the Advanced Test Reactor

The core of any nuclear reactor presents a particularly harsh environment for sensors and instrumentation. The reactor core also imposes challenging constraints on signal transmission from inside the reactor core to outside of the reactor vessel. In this paper, an acoustic measurement infrastructure installed at the Advanced Test Reactor (ATR), located at Idaho National Laboratory, is presented. The measurement infrastructure consists of ATR in-pile structural components, coolant, acoustic receivers, primary coolant pumps (PCPs), a data acquisition system, and signal-processing algorithms. Intrinsic and cyclic acoustic signals generated by the operation of the PCPs are collected and processed. Finally, the characteristics of the intrinsic signal can indicate the process state of the ATR (such as reactor startup, reactor criticality, reactor attaining maximum power, and reactor shutdown) during operation (i.e., real-time measurement).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Autonomous System Inference, Trojan, and Adversarial Reprogramming Attack and Defense (Final)

In the world of ever-advancing technology, Autonomous Systems (AS) find extensive application, bolstering functionalities of critical infrastructures such as nuclear power plants. These systems, however, are increasingly becoming a target for nefarious activities, namely through inference attacks, trojan attacks, and adversarial reprogramming. This paper delves into a comprehensive exploration of machine learning (ML)-driven autonomous control systems within advanced nuclear reactor designs, revealing the vulnerabilities and proposing strategies for defense against potential cyber-attacks. Advanced cyber-attacks against critical infrastructure and the energy sector are becoming more common. With the invention of autonomous control systems (ACS) within advanced nuclear reactor designs, system designers, reactor operators, and regulators must consider cybersecurity during the design and operational phases. This article provides a cyber threat assessment of machine learning (ML)- based digital twinning (DT) technologies in the context of advanced reactor ACS. A cyber-physical testbed was created to emulate nuclear reactor digital instrumentation and controls (I&C) and act as a basis for the ACS. The ACS was designed as two plant-level DTs predicting reactor malfunctions and determining control actions and two component-level DTs responsible for classifying component states and forecasting component inputs and outputs (I/O). Two duplicate ACS designs– one using a traditional ML framework and one using an automated ML (AutoML) framework– were created and tested against cyber-attacks on training data, real-time process data, and ML model architectures to determine their respective qualitative cyber-risk in terms of likelihood and impact. Both frameworks showed similar cyber-resilience against training, real-time, and ML architecture attacks, proving that neither is inherently more secure. Recommended safeguard and security measures are posed to system designers, reactor operators, and regulators to maintain the cybersecurity of ML-based DT technologies such as ACS, prompting a holistic view of shared responsibility for maintaining cyber-secure ML-based systems. As global reliance on generation III reactors begins to be critically assessed, the evolution towards advanced reactor systems utilizing digital instrumentation and controls (I&C) becomes not merely preferable, but essential. The integration of semi and fully autonomous control systems (ACS), powered by digital I&C and machine learning (ML)-based digital twinning (DT) technologies, emerges as a potent strategy to mitigate operations and maintenance costs, thereby enhancing the economic feasibility of novel reactor designs. However, with a staggering 500% and 380% increase in cyber-attacks reported against the energy sector by the United States Department of Energy (DoE) and the European Union respectively, a surge in cyber vulnerabilities specifically targeting the nuclear industry has been 2 markedly observed. Notable incidents, such as the W32.Ramnit spyware infiltration at the Gundremmingen nuclear power plant in Germany and the Dtrack spyware intrusion at the Kudankulam nuclear power plant in India, while not directly compromising core industrial control systems (ICS), underscore a compelling necessity to fortify cybersecurity protocols in safeguarding reactor systems against increasingly adept digital adversaries. In light of this, our investigation extends beyond conventional cybersecurity parameters, diving into the intricate web of potential vulnerabilities woven into ML-based DTs and ACS in advanced reactor systems. A crafted cyber-physical testbed and preliminary ACS were devised to act as a mirror, reflecting potential configurations of advanced reactor control designs. Moreover, this study is intertwined with a scrutinization of ML models, developed either through conventional, manually tuned methodologies or via automated means through AutoML, probing into their cyber-risk profiles within operational technology (OT) environments. Expanding on this, two distinct ACS blueprints were forged – one navigating through the corridors of traditional ML and the other traversing the path of AutoML – in an effort to holistically encapsulate the considerations pivotal to ML-based DT control system design. Employing the SANS Institute Industrial Control System (ICS) Kill Chain and the MITRE ATT&CK Tactics, Techniques, and Procedures (TTP) framework, a structured analysis was conducted, launching three targeted attacks against the training dataset, real-time dataset, and ML models, therein dissecting the potential cyber-attack implications against both ML frameworks within an ACS milieu. It is essential to note that three distinct categories of attacks were conducted against both ACS configurations, each encompassing three distinct ML-based DTs, cumulating in a total of 18 varied attacks. This exploration extends into the realms of Autonomous System Inference, Trojan, and Adversarial Reprogramming Attack and Defense, unraveling vulnerabilities, and opportunities for fortified defenses against such intrusions, particularly where ML-driven technologies, and by extension, ACS, are deployed. Final recommendations, articulated through a lens of security, safeguard, and implementation considerations, are presented for both traditional and AutoML models, anchoring upon the existing knowledge landscape and ML-based DT modeling for ACS, and are offered as a beacon to guide the nuclear industry through the intricate cybersecurity challenges that lie ahead.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Optimization of Silica Optical Fiber for High Temperature, Radiation Environments

The Department of Energy and Idaho National Laboratory are restarting the TREAT (Transient Reactor Test facility) reactor and are interested in updating the reactor's instrumentation regarding the analysis of the test fuel. Optical fibers' ability to make distributed temperature measurements coupled with its small footprint makes optical fiber a viable candidate to monitor fuel performance in the TREAT reactor. In addition to TREAT reactor, the U.S. Department of Energy and nuclear reactor companies like TerraPower have shown interest in using distributed optically based instrumentation for nuclear instrumentation in Next-Gen reactors. The work presented in this paper focuses on optimizing commercial silica optical fiber so that it can survive the temperatures and radiation fluences of the TREAT reactor and Next-Gen reactors. Temperatures in the TREAT reactor can spike to around 900 - 1000 C and so the goal of this work is to optimize optical fiber to survive these temperatures.

36 MATERIALS SCIENCE↗

Method for housing nuclear reactor modules

An in-core instrumentation system for a reactor module includes a plurality of in-core instruments connected to a containment vessel and a reactor pressure vessel at least partially located within the containment vessel. A reactor core is housed within a lower head that is removably attached to the reactor pressure vessel, and lower ends of the in-core instruments are located within the reactor core. The in-core instruments are configured such that the lower ends are concurrently removed from the reactor core as a result of removing the lower head from the reactor pressure vessel.

Keller, Michael↗

V and V techniques for FPGA-based I and C systems - How do they compare with techniques for microprocessors? - 355

We compare verification and validation (V and V) techniques for FPGA and microprocessor-based instrumentation and control (I and C) systems from the point of view of standards compliance, an approach based on behavioural properties, and the analysis of vulnerabilities. We found that the non-technology-specific elements of the standards considered are very similar. Differences are more marked when considering behavioural properties and vulnerabilities: the amount of effort required and confidence level obtained depend on a number of properties of the particular design under verification. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Industry analytics to optimize time based maintenance for I and C components - 363

Using Industry Analytics to optimize time based maintenance for I and C components allows a nuclear site to increase cost efficiency without impacting equipment reliability. Reductions in frequency or deletions for non-value based Preventive Maintenance (PM) improves cost efficiency by decreasing man-hours, parts costs, and planning costs of that PM; while maintenance performed too frequently can induce maintenance related failures. Benchmarking Industry PM Performance and Equipment Failure data allows sites to efficiently optimize time based maintenance frequencies. Equipment Failure history identifies gaps in maintenance frequencies as well maintenance-induced failures. Overlapping the PM frequency from a site with Industry PM Performance identifies opportunities to justify cost savings by reducing the PM frequency. The opportunities are then validated by the site's living PM program to ensure that the site can support the extension. This paper will discuss an Industry analytical technique used for maintenance frequency reductions. The major aspects evaluated under the technique are validation of the Industry data, review of the component's history and comparison of additional information. Industry data from 180 sites is used as a benchmark to identify opportunities for PM frequency extension on like-for-like components based on manufacturer/model and application. The component history composed of preventive maintenance feedback history and corrective maintenance history is reviewed to determine if the frequency can be extended without adversely affecting equipment reliability. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Toward biologically inspired self healing digital embedded devices: BIO-SymPle - 388

One of the more interesting concepts being postulated for next gen power plants, is the notion of highly resilient and survivable I and C systems. While resilience and survivability are 'cross system attributes', resilience and robustness will rely on highly robust and fault resilient devices as a foundation. This paper presents a new self-healing digital I and C architecture inspired by the operation of two biological concepts: human immune system and embryonic stem cells. The purpose of this research is look at different bio-inspired paradigms for organizing system functions so that they become inherently resilient due to device organization and behaviors. Our current proposed approach (called BIO-SymPle) is organized in two levels: critical service layer, which is providing the intended service of critical application, and healing layer, that supervises the operation of all repair mechanisms. To facilitate coordination, organized behavior among cells we employ network-based topology structures which is comprised of four active cells, four pre-generated redundant cells, and one embryonic stem cell. Collectively, the ensemble represents the building block that executes the local functionality based on the expression for DNA genetic codes stored inside each cell. The four bio-operational active cells are protected by their neighboring spare cells that imitate how white blood cells are distributed around the antigens in their locations. Ultimately, the proposed multi-layer architecture is able to cover a significant number of coincident faults and we believe architectures like BIO-SymPLe can positively impact the next generation digital I and C system. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Cyber threat assessment of machine learning driven autonomous control systems of nuclear power plants

We report advanced cyber-attacks against critical infrastructure and the energy sector are becoming more common. With the invention of autonomous control systems (ACS) within advanced nuclear reactor designs, system designers, reactor operators, and regulators must consider cybersecurity during the design and operational phases. This article provides a cyber threat assessment of machine learning (ML)-based digital twinning (DT) technologies in the context of advanced reactor ACS. A cyber–physical testbed was created to emulate nuclear reactor digital instrumentation and controls (I&C) and act as a basis for the ACS. The ACS was designed as two plant-level DTs predicting reactor malfunctions and determining control actions and two component-level DTs responsible for classifying component states and forecasting component inputs and outputs (I/O). Two duplicate ACS designs– one using a traditional ML framework and one using an automated ML (AutoML) framework– were created and tested against cyber-attacks on training data, real-time process data, and ML model architectures to determine their respective qualitative cyber-risk in terms of likelihood and impact. Both frameworks showed similar cyber-resilience against training, real-time, and ML architecture attacks, proving that neither is inherently more secure. Recommended safeguard and security measures are posed to system designers, reactor operators, and regulators to maintain the cybersecurity of ML-based DT technologies such as ACS, prompting a holistic view of shared responsibility for maintaining cyber-secure ML-based systems.

99 GENERAL AND MISCELLANEOUS↗

Analysis and improvement of input channel selection algorithm for reliability of control systems in nuclear power plants - 138

In most of non-safety NSSS (Nuclear Steam Supply System) control systems, a process input is composed of two channels for redundancy. In a normal situation, the average of both input channels is used in the control algorithm. If a failed channel is detected, the 'channel selection algorithm' chooses the remaining normal input channel. For a bumpless transition, many implemented control systems include the 'ramping rate' function, which makes it possible for the selected input to slowly approach the new input channel. However, it recently turned out that this function adopted for positive effects may result in unintended consequences for the related control system and the plant in certain cases. Due to a process transient in the plant according to a significant event, any related sensor input signal can rapidly change and finally exceed the pre-defined range for detecting an out-of-range event in a short time. During the process of channel conversion due to a rapid transient in the plant, the processed input from the redundant channels can be distorted if the ramping rate function is not properly implemented. As a result, it can result in a plant shutdown because the control system fails to perform the critical function to mitigate a plant transient. To solve these potential problems, we analyzed the overall causes and derived a proposed algorithm for improving reliability and performance of the control systems. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Magnetic and Mechanical Analysis of a Large Aperture 15 T Cable Test Facility Dipole Magnet

The US Department of Energy (DOE) Office of Science (SC), is funding a large bore “Cable Test Facility Magnet” for testing advanced cables and inserts in high transverse field. This is a joint effort between the Office of High Energy Physics (HEP) and the Office of Fusion Energy Sciences (FES). The background field magnet for this facility is being developed at Lawrence Berkeley National Laboratory (LBNL) while the cryostat and test facility will be located and operated by Fermi National Accelerator Laboratory (FNAL). The Nb 3 Sn dipole magnet, which will provide the transverse background field, is designed to generate a field of 15 T in a 100 × 150 mm bore at 1.9 K. The conceptual design of a block-type dipole with flared ends and a structure based on key-and-bladder technology will be introduced. The results of the magnetic and mechanical analysis will be presented.

43 PARTICLE ACCELERATORS↗

A next generation ultra short pulse reflectometry (USPR) diagnostic

Ultrashort Pulse Reflectometry (USPR) is a plasma diagnostic technique involving the propagation and reflection of ultrashort duration (~few ns) chirps. The reflected packets pass through a multichannel filter with time-of-flight measurements performed on each of the filtered packets. A next generation USPR system is under development, spanning 28–75 GHz, for use on compact, short duration, magnetically confined fusion devices. This system presents a dramatic increase in performance compared with an earlier USPR system employed on the LLNL Sustained Spheromak Physics Experiment device more than a decade ago. The new system replaces upconverting mixers with higher power active multiplier chains to generate mm-wave transmitter chirps, with custom time-of-flight electronics reducing the time per measurement by a factor of 3X. Finally, the system is equipped with a field programmable gate array for data acquisition and analysis.

47 OTHER INSTRUMENTATION↗

Efficient loading of reduced data ensembles produced at ORNL SNS/HFIR neutron time-of-flight facilities

We present algorithmic improvements to the loading operations of certain reduced data ensembles produced from neutron scattering experiments at Oak Ridge National Laboratory (ORNL) facilities. Ensembles from multiple measurements are required to cover a wide range of the phase space of a sample material of interest. They are stored using the standard NeXus schema on individual HDF5 files. This makes it a scalability challenge, as the number of experiments stored increases in a single ensemble file. The present work follows up on our previous efforts on data management algorithms, to address identified input output (I/O) bottlenecks in Mantid, an open-source data analysis framework used across several neutron science facilities around the world. We reuse an in-memory binary-tree metadata index that resembles data access patterns, to provide a scalable search and extraction mechanism. In addition, several memory operations are refactored and optimized for the current common use cases, ranging most frequently from 10 to 180, and up to 360 separate measurement configurations. Results from this work show consistent speed ups in wall-clock time on the Mantid LoadMD routine, ranging from 19% to 23% on average, on ORNL production computing systems. The latter depends on the complexity of the targeted instrument-specific data and the system I/O and compute variability for the shared computational resources available to users of ORNL’s Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR) instruments. Nevertheless, we continue to highlight the need for more research to address reduction challenges as experimental data volumes, user time and processing costs increase.

Godoy, William↗

Modelling I and C systems with relays for simulation, adapting CIM standards - 243

As the nuclear power plants are in operation for many years, a main issue is to integrate simulation activities in parallel to the present I and C design process in order to limit impacts on current operating I and C engineering process and commissioning activities. The objective is to keep the existing I and C data and diagrams format. EDF chose to build an intermediate file independent of the tool used for simulation to exchange and store data. This file has been created adapting CIM standards (IEC 61968-13 and IEC 61970-301). It is an XML file built according to an RDF schema. CIM standards are intended to model transport or distribution electrical networks and used to facilitate system integration and information exchanges. We chose to adapt CIM to model EDF electromechanical relay circuitries in order to re-use all methods and tools developed around CIM standards such as RDF schema and RiseClipse, an open source tool developed by EDF and SUPELEC engineering school that will be used to verify exported I and C data according to a data model and to EDF engineering rules. This paper will describe the ongoing project on simulation in parallel to current I and C design process, focusing on how I and C electromechanical relays data are modelled based on CIM standards, and what benefits that model provides to I and C Verification and Validation activities. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Graded approach for assessing digital system failure susceptibilities - 305

Owner/operators of nuclear plants must be able to identify and assess susceptibilities to digital system failures and unintended behaviors that could lead to plant system malfunctions, including common-cause failures (CCFs) of multiple controlled components that may impact overall plant safety. Nuclear plant designers and regulators often assess and manage potential failure modes by assuming the failure occurs and showing by analysis that the results are acceptable. In 2016 EPRI published a guideline on Methods for Assuring Safety and Dependability when Applying Digital Instrumentation and Control Systems (EPRI 3002005326), which takes a more holistic approach that considers digital system failure modes from the perspective of their impact on plant risk and includes a graded approach based on safety significance. This paper describes that graded approach. The methodology in EPRI 3002005326 systematically identifies potential I and C vulnerabilities that could lead to significant malfunctions of controlled components and systems, including common-cause failures (CCF), and discusses in detail methods to protect against them. It considers both preventive measures that reduce the likelihood of failures, and plant systems and features that mitigate the effects of component failures and misbehaviors. Coping analysis is then performed as appropriate to provide additional assurance of protection. The safety significance based graded approach is an important feature in that it can help users focus attention on overall plant safety, including potentially risk-significant scenarios that might not be considered in traditional safety analyses. The purpose is to help the user ensure that modifications to the I and C that could potentially be safety significant are treated appropriately and at the same time not waste valuable resources on excessive protection against changes that have little or no impact on safety. The graded approach allows the design engineer to tailor the rigor of the preventive, limiting, and mitigative measures commensurately with the effects the I and C modification has on overall plant safety. The graded approach focuses on safety significance impact, which is effectively a qualitative measure of the potential change in safety (or risk) caused by a proposed I and C modification, as compared to the I and C that is being replaced. Safety significance impact should not be confused with importance to safety or risk significance. It is possible that an upgrade to safety-related and/or risk-significant I and C systems may have little or no safety significant impact and vice versa. The approach considers three factors that influence the impact on safety significance of a proposed I and C modification: context, likelihood of failure, and consequences of failure. The most significant of the three factors is context - what the I and C is connected to, both directly and indirectly, including its potential effects on plant systems that respond to transients or accidents. The graded approach is particularly helpful in assessing potential CCFs resulting from digital I and C failures, and shows why the most problematic CCFs will be those that can affect multiple plant systems. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

Procedure design of F and B operation and its impact on the risk of a nuclear power plant - 168

A feed and bleed (F and B) operation is an operator's emergency task to remove core heat from a reactor coolant system by bleeding coolant from the pressurizer, and feeding makeup water through safety injection systems under the loss of all feedwater condition. Even though the F and B operation is an important task having a significant impact on the risk of a nuclear power plant, there is a high uncertainty regarding its modeling and analysis in a PSA. This paper introduces a study on the design of an operational strategy for an F and B operation and its impact on the risk of a nuclear power plant in Korea. An intensive support analysis was conducted to understand the plant behavior and revise the success criteria for the HRA of the F and B operation. A human performance analysis on the emergency tasks was also conducted using a training simulator to generate the operator's performance time for the F and B operation. We revised the PSA model and updated the risk of the reference plant, which shows that the risk of the revised model is much higher than that of the existing model. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

On-Line Monitoring to Detect Sensor and Process Degradation under Normal Operational Transients - 327

Calibration assessment of nuclear plant sensors and transmitters is currently performed through periodic maintenance actions during plant refueling outages. On-line monitoring (OLM) can be used to reduce the time and cost of this maintenance process by using empirical models to evaluate sensor calibration. The implementation of OLM can also mitigate human error associated with unnecessary maintenance on sensors operating within calibration tolerances. OLM has been studied and validated for plants under steady-state baseload operation, but the extension to load-following operation is not well-documented in the literature. Auto-associative kernel regression (AAKR) provides an error-corrected prediction of process sensor measurements for OLM through a weighted average of historic data. The current study evaluates the performance of AAKR for prediction under normal operational transients (such as expected under load-following operation). Performance of the AAKR model with two anomaly detection routines, error uncertainty limit monitoring (EULM) and sequential probability ratio test (SPRT), was evaluated for normal operation, sensor faults, and process degradation. The data used to develop and evaluate the OLM system were collected from a forced flow loop with temperature and flow sensors. Three models of well-correlated groups of sensors are trained to predict normal system behavior under planned transients. The anomaly detection routines are tested for sensor faults and process faults. The results of this experimental study suggest AAKR and an appropriate anomaly detection routine can be applied for sensor calibration and process fault detection under transient operation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Methodological approach to the sensitivity analysis of failure effects in modern digital I and C systems - 136

Modern I and C systems of nuclear power plants increasingly use digital equipment, software-based applications for I and C functions and different types and topologies of communication networks as well. The structure, operation and communication of digital safety-related I and C systems are strongly influenced by the implemented automatic fault detection and fault treatment procedures as barriers against failure propagation. Thus, these I and C systems can change their operating states automatically during the operation time, e.g. due to the detection and correction of certain failure types. Reasonably because of these dynamic properties of digital I and C systems the use of classical fault tree and event tree methods solely for the reliability analysis of modern I and C systems is less suitable. These methods basically apply static models and do not consider the temporal changes of the operating state of the I and C system or equipment appropriately. In this paper, a recently developed approach for the sensitivity analysis of digital I and C systems will be introduced that allows to investigate the dynamic behavior of such systems in cases of internal faults and to quantify the effects on the system reliability. The development and testing of the analysis methodology is carried out using models of generic I and C systems with different typical architectures of stepwise increased complexity. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗