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At least 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Certification of RadICS FPGA-based platform under US NRC requirements - 386

The paper discusses approach, applicable to certify digital instrumentation and control (I and C) platform under US Nuclear Regulatory Commission (NRC) requirements. Currently, Radics LLC is certifying its Field-Programmable Gate Array (FPGA)-based RadICS Platform using Platform-based test specimen. The Platform has been recently certified under IEC 61508 to meet Safety Integrity Level (SIL) 3 requirements. Up to now Radics LLC has developed a number of FPGA-based I and C applications based on RadICS Platform. Those applications include applications based on customer's specification and reverse engineering applications (Fit, Form and Function reproduction of existing equipment). The life cycles of applications include independent verification and validation of FPGA-based I and C systems/components, Commercial Grade Dedication of components for FPGA-based I and C systems, turnkey implementation of I and C project, supply of hardware, engineering tools and application design process. The paper also describes in details certification approach, which is based on Digital I and C-ISG- 06 and includes the following stages: general agreement with the US NRC, Topical Report submittal, timely responses and successful audits, and timely issuance of approved Topical Report. Such approach is supported by 10 CFR 50 Appendix B Quality Management System, which is also discussed in the paper. In addition, an overview of RadICS Platform Topical Report is presented, including its overview and structure, certification scope and obligatory appendixes. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Development of a bayesian belief network model for the software reliability assessment of nuclear digital I and C safety systems - 206

Since the digital instrumentation and control systems are expected to play an important role for the safety systems in nuclear power plants (NPPs), the need has emerged to not only establish a basis for incorporating software behavior into digital I and C system reliability models, but also to quantify the failure probability of the software used in NPP digital protection systems. In this study, a Bayesian belief network (BBN) model is developed to quantitatively assess software reliability by estimating the number of faults in a software program considering its software development life cycle (SDLC). The model structure and parameters are established based on the information applicable to NPP safety-related systems and the evidence used to construct and quantify the BBN model was collected from three stages of expert elicitation. The software failure probability is estimated from the number of residual defects in a software program at the end of SDLC phase. As a case study, the BBN model was applied to quantify the software reliability of a typical digital protection software having the size of 50 function points and having the Medium development and validation and verification (V and V) qualities. The developed model can be applied to estimate the failure probability for both developing and deployed safety-related NPP software, and such results can be used to evaluate the quality of the digital I and C systems in addition to estimating potential reactor risk due to software failure. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Analysis of, and defense against, spurious actuations and other incorrect behaviors of digital control systems - 390

Modern digital instrumentation and control (I and C) systems offer nearly unlimited functional capability. However, the complexity of their design (including their system software and application software) is significantly higher than that of conventional, hardwired systems, to the point that it is generally impossible to demonstrate that there are no residual design errors, even when the best development and verification techniques have been applied. When activated by specific conditions, design errors could result in systematic failures, i.e., failures that occur deterministically each time and wherever an activating condition is present. Redundancy could be defeated when the same design error affects multiple channels or systems, and the activating condition affects several of these channels or systems. In the case of safety I and C systems, most of the functions are poised protection functions that must be actuated when and only when a demand situation occurs in the plant. In nuclear power plants, the most dangerous failure mode for such functions generally is failure to actuate, and many measures are taken to protect from such failure mode. This paper considers the other failure modes of safety I and C functions, and also the failure modes of the other I and C functions that are less important to safety. It is not limited to poised functions, and considers other types of functions such as regulation functions or human interface functions. It proposes a framework for analyzing and providing defense against spurious actuations and other incorrect behaviors from digital I and C systems directly interacting with the plant process or with operators: these might be useful for example to justify that design bases for reactor protection are adequate, or to ensure plant performance and availability. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Secure environment establishment for FPGA-based safety-critical systems: quality management system context - 159

The paper discusses importance of secure development and operational environment establishment process, its particular stages, including security-oriented analysis and assessment of safety-critical instrumentation and control (I and C) systems. Two particular aspects are covered - security assessment and security assurance related to processes established by Quality Management System (QMS). One of the underlying ideas for the security assessment approach is in performance of gap analysis. Such analysis considers influence of all different factors within I and C system development process, and further application of Intrusion Modes and Effect Criticality Analysis (IMECA) technique. It makes the approach applicable to various safety-critical systems, considering underlying technologies (for example, Field-Programmable Gate Arrays (FPGA). Nowadays FPGA technology is often used in wide spectrum of hardware devices - from personal devices to complex I and C systems for Nuclear Power Plants (NPPs). The problem of cyber security assessment and assurance for FPGA technology in general and its application in safety-critical I and C systems in particular is not completely solved due to several reasons that cover regulations, QMS processes, implementation of development and verification activities, etc. The paper overviews existing regulations, which cover various aspects of NPP I and C systems development and operation, FPGA technology implementation, and secure development and operational environment establishment and assurance. The paper also emphasizes that complex approach should be used to identify and mitigate all specific vulnerabilities thorough the whole life cycle on the basis on Intrusion Modes, Effect and Criticality Analysis technique. The paper also proposes approach to the development of safety-critical I and C systems in the scope of cyber security assurance process at developer site. Secure development environment is described as a key factor during the development process and the paper also discusses establishment and maintenance of such environment as a part of company's Nuclear Quality Assurance's QMS. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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 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↗