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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 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

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↗
DOE OSTI2022

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↗
DOE OSTI2014

Status Report on Efforts to Enhance Instrumentation to Support Advanced Test Reactor Irradiations

The Department of Energy (DOE) designated the Advanced Test Reactor (ATR) as a National Scientific User Facility (NSUF) in April 2007 to support the growth of nuclear science and technology in the United States (US). By attracting new research users - universities, laboratories, and industry - the ATR NSUF facilitates basic and applied nuclear research and development, further advancing the nation's energy security needs. A key component of the ATR NSUF effort at the Idaho National Laboratory (INL) is to design, develop, and deploy new in-pile instrumentation techniques that are capable of providing real-time measurements of key parameters during irradiation. To address this need, an assessment of instrumentation available and under-development at other test reactors was completed. Based on this initial review, recommendations were made with respect to what instrumentation is needed at the ATR, and a strategy was developed for obtaining these sensors. In 2009, a report was issued documenting this program’s strategy and initial progress toward accomplishing program objectives. Since 2009, annual reports have been issued to provide updates on the program strategy and the progress made on implementing the strategy. This report provides an update reflecting progress as of January 2014.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

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 ↗
DOE OSTI2017

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↗
DOE OSTI2017

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↗
DOE OSTI2016

PREDICTIVE MODELING OF ACOUSTIC SIGNALS FROM THERMOACOUSTIC POWER SENSORS (TAPS)

Thermoacoustic Power Sensor (TAPS) technology offers the potential for self-powered, wireless measurement of nuclear reactor core operating conditions. TAPS are based on thermoacoustic engines, which harness thermal energy from fission reactions to generate acoustic waves by virtue of gas motion through a porous stack of thermally nonconductive material. TAPS can be placed in the core, where they generate acoustic waves whose frequency and amplitude are proportional to the local temperature and radiation flux, respectively. TAPS acoustic signals are not measured directly at the TAPS; rather, they propagate wirelessly from an individual TAPS through the reactor, and ultimately to a low-power receiver network on the vessel’s exterior. In order to rely on TAPS as primary instrumentation, reactor-specific models which account for geometric/acoustic complexities in the signal propagation environment must be used to predict the amplitude and frequency of TAPS signals at receiver locations. The reactor state may then be derived by comparing receiver signals to the reference levels established by predictive modeling. In this paper, we develop and experimentally benchmark a methodology for predictive modeling of the signals generated by a TAPS system, with the intent of subsequently extending these efforts to modeling of TAPS in a liquid sodium environmen

thermoacoustics, wireless in-core radiation detect↗
DOE OSTI2023

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↗
DOE OSTI2013

Selecting a radiation tolerant piezoelectric material for nuclear reactor applications

Bringing systems for online monitoring of nuclear reactors to fruition has been delayed by the lack of suitable ultrasonic sensors. Recent work has demonstrated the capability of an AlN sensor to perform ultrasonic evaluation in an actual nuclear reactor. Although the AlN demonstrated sustainability, no loss in signal amplitude and d{sub 33} up to a fast and thermal neutron fluence of 1.85 Multiplication-Sign 1018 n/cm{sup 2} and 5.8 Multiplication-Sign 1018 n/cm{sup 2} respectively, no formal process to selecting a suitable sensor material was made. It would be ideal to use first principles approaches to somehow reduce each candidate piezoelectric material to a simple ranking showing directly which materials one should expect to be most radiation tolerant. However, the complexity of the problem makes such a ranking impractical and one must appeal to experimental observations. This should not be of any surprise to one whom is familiar with material science as most material properties are obtained in this manner. Therefore, this work adopts a similar approach, the mechanisms affecting radiation tolerance are discussed and a good engineering sense is used for material qualification of the candidate piezoelectric materials.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

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↗
DOE OSTI2021

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↗
DOE OSTI2021

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↗
DOE OSTI2021

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↗
DOE OSTI2017

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↗
DOE OSTI2016

Common Cause Failure Mitigation Experience on Digital I and C System in NPP

CCF is failures of redundant or separate structures, systems, or components due to a shared root cause. The CCF includes those are resulted from severe environments, design errors, calibration and maintenance errors, and consequential failures. In addition, digital I and C systems can be vulnerable to CCF caused by software errors or software developed logic, which could defeat the redundancy achieved by the hardware architecture. Therefore, diversity and defense in-depth (D3) has become an important issue to overcome the CCF of the safety I and C system. The U.S. NRC has established regulatory guidance addressing a method for assessing D3 provided by the safety I and C system architecture at a NPP. This method enables determination of whether vulnerabilities to CCF have been adequately addressed. The guidance is included in BTP 7-19 and Chapter 7 of NUREG-0800. The regulatory guidance provides a methodology for designing diversity. APR1400 establishes the mitigating strategies that consider the characteristics of digital I and C system that contribute to the potential for CCF. The diversity strategies consist of a combination of diversity attributes and associated criteria to address potential CCF vulnerabilities in digital safety I and C system. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Suggestion of a Framework to Evaluate the Extent of Cyber Security Enhancements in NPP

Digital I and C systems have been developed and installed in nuclear power plants (NPPs). However, due to installation of digital I and C systems, cyber security concerns have been emerged. One of the major problems is that nuclear industry is in very early stage in dealing with cyber security issues. It is because that cyber security has received less attention compared to other safety problems. In addition, late adoption of digital I and C systems has resulted in lower level of cyber security advancements in nuclear industry than ones in other industries. With this regards, recently, international standard organizations, regulatory institutes, and research institutes have performed a number of studies addressing these cyber security issues in nuclear industries. In order to provide useful information about cyber security issues including cyber security enhancements, this paper suggests a framework to evaluate how much cyber security is improved when a specific cyber security enhancement is applied in NPPs. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

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 ↗