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At least 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Initial Findings in Qualifying the Instrument Thimble 11 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 AN↗
DOE OSTI2017

Development of a systematic approach for safety assessment of PLD based equipment of I and C systems - 354

Nuclear power plants (NPPs) increasingly apply digital equipment for instrumentation and control (I and C) systems. During the life cycle of NPPs the operators, designers and manufactures are facing the need to replace or upgrade obsolete equipment or to implement new digital I and C systems. A relatively new challenge in this area is the deployment of programmable logic devices (PLDs) for the equipment of I and C systems. Regulatory approval for safety and safety related I and C systems demands confirmation of evidence that design as well as verification and validation (V and V) processes have ensured that the application of new equipment will fulfil safety and reliability requirements and its intended design function. The objective of the recent research project of GRS is developing a systematic approach for the assessment of PLD-based equipment in safety and safety-related I and C systems regarding the fulfilment of regulatory requirements and identification of undesired failure effects. A generic approach for the safety-oriented assessment of PLD-based equipment has been developed, based on the principles of a Plan-Do-Check-Act (PDCA) cycle. The approach comprises planning of the assessment, predefined analysis steps and methodologies, as well as required input information, criteria for decision making process, the tasks concerning identification of improvement potential and determining documentation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Systems and methods for processing irradiation targets through a nuclear reactor

Apparatuses and methods produce radioisotopes in instrumentation tubes of operating commercial nuclear reactors. Irradiation targets may be inserted and removed from instrumentation tubes during operation and converted to radioisotopes otherwise unavailable during operation of commercial nuclear reactors. Example apparatuses may continuously insert, remove, and store irradiation targets to be converted to useable radioisotopes or other desired materials at several different origin and termination points accessible outside an access barrier such as a containment building, drywell wall, or other access restriction preventing access to instrumentation tubes during operation of the nuclear plant.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2013

Simplified modeling of liquid sodium medium with temperature and velocity gradient using real thermal-hydraulic data. Application to ultrasonic thermometry in sodium fast reactor

In the framework of the French R and D program for the Generation IV reactors and specifically for the sodium cooled fast reactors (SFR), studies are carried out on innovative instrumentation methods in order to improve safety and to simplify the monitoring of fundamental physical parameters during reactor operation. The aim of the present work is to develop an acoustic thermometry method to follow up the sodium temperature at the outlet of subassemblies. The medium is a turbulent flow of liquid sodium at 550 Degree-Sign C with temperature inhomogeneities. To understand the effect of disturbance created by this medium, numerical simulations are proposed. A ray tracing code has been developed with Matlab Copyright-Sign in order to predict acoustic paths in this medium. This complex medium is accurately described by thermal-hydraulic data which are issued from a simulation of a real experiment in Japan. The analysis of these results allows understanding the effects of medium inhomogeneities on the further thermometric acoustic measurement.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

The key-role of instrumentation for the new generation of research reactors

Experimental reactors have been indispensable since the beginning of the use of nuclear energy to support many important fields of industry and research: safety, lifetime management and operation optimisation of nuclear power plants, development of new types of reactors with improved resources and fuel cycle management, medical applications, material development for fusion... Over the last decade, modifications of the operational needs and the ageing of the nuclear facilities have led to several closures and time is coming for new key European Experimental Reactors (EER) within a European and International Framework. Projects like MYRRHA and JHR are underway to define and implement a new consistent EER policy: - Meeting industry and public needs, keeping a high level of scientific expertise; - With a limited number of EER, specified within a rational compromise between specialisation, complementarities and back-up capacities; - To be put into effective operation in this or the next decade. These new projects will give to the scientific community high performances allowing innovative fields of R and D. A new generation of instrumentation to address new phenomena and that allows better on-line investigation of some key physical parameters is necessary to achieve these challenges. One initiative to progress in this direction is the Joint Instrumentation Laboratory between CEA and SCK.CEN which has already given significant results and patents. Major scientific challenges to achieve in the field of instrumentation for this new generation of European Research Reactors have to be investigated and are described in this paper as well as a short description of the JHR and MYRRHA reactors that will be serving as flexible irradiation facilities for testing them. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Development of a new IEC Standard on cybersecurity controls for I and C in Nuclear Power Plants - IEC 63096 - 158

Standards are promoting the enhancement of cybersecurity in Nuclear Power Plants (NPP). Standardizing processes and procedures are also crucial to achieve successful international collaboration. Many information security standards were developed in recent years; however, not many of them could be directly applied to NPPs. This paper provides an overview of the development of a new cybersecurity standard dealing with cybersecurity controls for nuclear power plants around the world. For I and C systems in NPPs, the new IEC 63096 standard targeted for issue in 2019, focuses specifically on the selection and application of cybersecurity controls from the included security controls catalogue. The role of selected and applied cybersecurity controls is to prevent, detect and react to digital attacks against computer-based I and C systems. This standard applies to the I and C of new nuclear power plants and to the I and C modernization in existing plants. It is being prepared and based on IEC 62645 , ISO/IEC 27000 series , IAEA , and country specific guidance in this expanding the technical and the security focus area. The Standard is intended to be used by designers and operators of NPPs (utilities), systems evaluators, vendors and subcontractors, and by licensors. While the development of international standards is essential it is also a lengthy process, taking multiple years. Especially with regard to cybersecurity, all involved stakeholders have to collaborate. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 OSTI2012

I&C Modeling in SPAR Models

The Standardized Plant Analysis Risk (SPAR) models for the U.S. commercial nuclear power plants currently have very limited instrumentation and control (I&C) modeling [1]. Most of the I&C components in the operating plant SPAR models are related to the reactor protection system. This was identified as a finding during the industry peer review of SPAR models. While the Emergency Safeguard Features (ESF) actuation and control system was incorporated into the Peach Bottom Unit 2 SPAR model in a recent effort [2], various approaches to expend resources for detailed I&C modeling in other SPAR models are investigated.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗