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At least 91 records · Page 5
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Demonstrating and argumenting safety of I and C systems - Challenges and recent experiences - 180

Nuclear licensees are required to submit a documented justification of the safety of the plant and its systems to the local regulatory body. Developing this documentation is a hard task and requires a lot of effort from several stakeholders. It should be understandable, transparent, convincing and complete. Observations from the current practice indicate that the nuclear power industry would benefit from more structured, systematic and harmonized practices in engineering instrumentation and control (I and C) systems and justifying their safety. This paper describes recently recognized challenges in demonstrating the safety of digital I and C systems and suggests possible ways to solve them. Some of the proposed solutions are light improvements in the everyday documentation practices, working methods and utilization of computerized support tools, while others would involve fundamental changes in the design and documentation philosophy that are more demanding to implement. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
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 OSTI2017

Risk informed cyber security for nuclear power plants - 342

Nuclear power plants are increasingly adding digital components for plant operation, safety, and security. These digital components fill a gap with legacy equipment where replacement components no longer exist. They also benefit operation of the plant by increasing efficiency in power generation, monitoring of equipment and plant parameters, as well as aiding operator control. However, the addition of digital components and systems also adds cyber risks with previously unanalyzed failure modes and attack vectors are introduced with these new systems. These risks are difficult to identify, analyze, and mitigate due to the increasingly complex nature of the digital components and the integration of these components with additional plant processes and communication networks. The research presented in this paper develops a new method that addresses the cyber risk to inform appropriate levels of protection. EPRI and Sandia are working under a Cooperative Research and Development Agreement to develop an effective method of evaluating the cyber risk in production nuclear power facilities. The Cyber Hazards Analysis Risk Methodology (CHARM) focuses on ensuring adequate controls are in place for appropriate cyber protection of the plant from radiological release or generation risk. Existing plant hazards analyses (e.g., PRA, FTA) do not account for software deficiencies or adversarial intent. This method leverages existing plant analyses and MIT's Systems Theoretic Process Analysis (STPA) to create cyber informed fault trees. These new fault trees will provide the basis for comprehensive cyber risk analysis and help ensure potential gaps in cyber security controls are identified and corrected. (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

Automated work package: capabilities of the future - 275

As part of the ongoing efforts at the U.S. Department of Energy's Light Water Reactor Sustainability Program, Idaho National Laboratory (INL) is conducting several pilot projects in collaboration with the nuclear industry to improve the reliability, safety, and economics of the nuclear power industry, especially as the nuclear power plants extend their operating licenses to 80 years. One of these pilot projects is the automated work package (AWP) pilot project. An AWP is an electronic intelligent and interactive work package. It uses plant condition, resources status, and user progress to adaptively drive the work process in a manner that increases efficiency while reducing human error. To achieve this mission, the AWP acquires information from various systems of a nuclear power plant and incorporates several advanced instrumentation and control technologies along with modern human factors techniques. With the current rapid technological advancement, it is possible to envision several available or soon-to-be-available capabilities that can play a significant role in improving the work package process. In this pilot project, the researchers develop an AWP prototype of an expanding set of capabilities and evaluates them in an industrial environment. While some of the proposed capabilities are based on using technological advances in other applications, others are conceptual; thus, they require significant research and development to be applicable in an AWP. The scope of this paper is to introduce a set of envisioned capabilities, their need for the industry, and the industry difficulties they resolve. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Application of a bayesian belief network model to reliability assessment of nuclear safety-related software - 207

As the instrumentation and control (I and C) systems in nuclear power plants (NPPs) have been replaced with digital-based systems, the need to incorporate software failures into NPP probabilistic risk assessments has arisen. In order to assess the probability of software failure on demand, a Bayesian belief network (BBN) model was developed which estimates the number of defects and the resulting probability of software failure on demand in nuclear safety-related software. To assess the feasibility of the BBN framework, the BBN model was applied to the prototype Integrated Digital Protection System-Reactor Protection System (IDiPS-RPS) to estimate the number of remaining faults and the software failure probability of a target software. The developmental- and V and V-activities carried out during the IDiPS-RPS development process were evaluated based on the well-defined checklist derived by the V and V team and were estimated based on expert elicitation. In addition, the attribute evaluations and the number of FPs of the target software is provided as the inputs for the BBN model. The application results showed the feasibility of using BBNs for quantifying software failure probabilities and several insights were gained from the applications of the BBN model. The proposed BBN framework can be applied to estimate the software failure probability for other safety-related NPP software and provide an insight on modeling the software development process that involves iterations between different development phases. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Griffin Capability Improvements in Support of Ex-core Deep-Penetration Problems

Advanced reactor designs, especially portable reactors that are designed to be located closer to humans and operate autonomously, require the ability to accurately compute the ex-core neutron and gamma flux solutions in terms of shielding design optimization to reduce dose rates at the vessel boundary and detector signal prediction to drive the reactor control system. The Nuclear Energy Advanced Modeling and Simulation program has prioritized improvements to the Griffin discrete ordinates (SN) solver for deep-penetration problems in fiscal year 2025. Significant advancements have been made to the Griffin methodologies for solving ex-core deep-penetration problems for steady-state, fixed-source and transient calculations. This work presents the methodology improvements as well as a comprehensive demonstration with a Transient Test Reactor model and measurements.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
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

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↗
NASA NTRS1959

NASA Reactor Facility Hazards Summary

The Lewis Research Center of the National Aeronautics and Space Administration proposes to build a nuclear research reactor which will be located in the Plum Brook Ordnance Works near Sandusky, Ohio. The purpose of this report is to inform the Advisory Committee on Reactor Safeguards of the U. S. Atomic Energy Commission in regard to the design Lq of the reactor facility, the characteristics of the site, and the hazards of operation at this location. The purpose of this research reactor is to make pumped loop studies of aircraft reactor fuel elements and other reactor components, radiation effects studies on aircraft reactor materials and equipment, shielding studies, and nuclear and solid state physics experiments. The reactor is light water cooled and moderated of the MTR-type with a primary beryllium reflector and a secondary water reflector. The core initially will be a 3 by 9 array of MTR-type fuel elements and is designed for operation up to a power of 60 megawatts. The reactor facility is described in general terms. This is followed by a discussion of the nuclear characteristics and performance of the reactor. Then details of the reactor control system are discussed. A summary of the site characteristics is then presented followed by a discussion of the larger type of experiments which may eventually be operated in this facility. The considerations for normal operation are concluded with a proposed method of handling fuel elements and radioactive wastes. The potential hazards involved with failures or malfunctions of this facility are considered in some detail. These are examined first from the standpoint of preventing them or minimizing their effects and second from the standpoint of what effect they might have on the reactor facility staff and the surrounding population. The most essential feature of the design for location at the proposed site is containment of the maximum credible accident.

Source record↗
DOE OSTI2017

Continuous measurement of boron concentration in nuclear power plants: why is it essential for safe and efficient operation? - 318

As a neutron-absorber, Boron 10 plays a significant role in nuclear power plants as it is used in order to control reactivity in the core, together with control rods. For that reason; it is a key parameter that has to be controlled. In a similar way, it is crucial to be able to detect a boron dilution in the primary circuit, as it would causes severe damages if not detected and controlled. Thus, it comes as evidence that a boron concentration monitoring system is a key for every nuclear power plant. Consequently, it becomes important to provide plant operators with a boron concentration measurement issued by a reliable, redundant, classified and qualified system, with high accuracy and short response time. In order to cover this need, Rolls-Royce has designed Boronline{sup R}, the latest generation of boron meter, which provides nuclear power plant operators with a permanent record of boron concentration of primary circuit by way of continuous sampling and real-time measurements. Boronline takes part of a new generation of Advanced Sensors and Measurements Technologies. In this paper, boron meter role, main functions and measuring principles will be detailed, as well as technical characteristics of the new generation of boron meter will be presented. (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 OSTI2017

Complexity and error potential of digital I and C - 135

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

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE 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 OSTI2020

ARCS Replacement at TREAT for TRTR 2020 - Abstract

The Transient REActor Test (TREAT) facility was restarted in 2017 to perform reactor fuels testing under transient conditions, following shutdown in 1994 and 23 years in non-operational standby. TREAT is a transient reactor that, like most pulsing reactors, can create natural shaped power pulses. However, TREAT provides the ability to shape the power pulse and control the energy deposition rate on the test specimens by automatically controlling the amount of reactivity insertion over time. The original automatic reactor control system (ARCS) was created in the 1980s using INTEL 8086 CPU boards and software written in assembly code. The need to maintain the system for the next 40+ years of TREAT life led to the decision to replace ARCS with modern hardware while maintaining the original functionality. After a prototyping effort, the replacement was successfully completed in the first quarter of 2020. The process and results of that effort are reported here.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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 OSTI2017

Qualified display system architecture - 246

There is a need in the nuclear industry for a Qualified Display System (QDS) architecture as described in this paper. The function of the QDS is to process and display Regulatory Guide 1.97 post accident monitoring parameters. The QDS is implemented on the Lockheed Martin Field-Programmable Gate Array (FPGA) based Nuclear Protection and Control (NuPAC) digital platform, coupled with a Qualified Display Controller (QDC) and a Safety Display Unit (SDU). The QDS is designed to meet the requirements of IEEE 603-1991 (endorsed by Regulatory Guide 1.153, Rev. 1), IEEE 497-2002 (endorsed by Regulatory Guide 1.97, Rev. 4), and IEEE 7-4.3.2- 2003 (endorsed by Regulatory Guide 1.152, Rev. 3). The QDS is designed with flexibility such that it can be implemented in an advanced reactor design, Small Modular Reactor (SMR) with any cooling medium, or digital upgrade project for an operating nuclear plant. The QDS has four independent and redundant divisions. Each division has five layers in the architecture: Process variable input signals; Process Protection Logic (PPL); Algorithm Logic (AL), Interface Panel (IP); and Safety Display Unit (SDU). The QDS meets the following design criteria: redundancy, independence, predictability and repeatability, and diversity capability. The QDS architecture also meets the simplicity design principle. The QDS architecture design provides flexibility in the implementation in a plant. The QDS can be implemented in the design of a new advanced plant or as a digital upgrade in an operating plant. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Practical applications of model checking in the finnish nuclear industry - 281

Model checking is a powerful, formal, computer-assisted verification method that can be used to prove that a model of a (hardware or software) system fulfills stated properties. When used right, model checking can prove the correctness of instrumentation and control (I and C) system application logic, be it a software or a field-programmable gate array (FPGA) based design. The verified properties can also address unwanted functionality, making model checking a very effective method for analyzing spurious actuation. Despite the benefits, model checking is still not widely adopted in the verification of I and C systems, one exception being the Finnish nuclear industry. Since 2008, VTT has applied model checking in practical customer work related to the Olkiluoto 3 EPR, Loviisa 1 and 2 VVER-440, and Hanhikivi 1 AES-2006 nuclear power plants, on commission from either the regulator or the utilities. In this paper, we look at how the method has been used in each of these three projects. We also introduce a user-friendly, graphical modelling tool called MODCHK. We then present an overall view of the design issues detected during nine years of customer work, and one practical example. The 100% coverage means that the design errors that are found often relate to scenarios that are otherwise hard to account for (e.g., exact timing of events, or improper operator actions). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗