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All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Risk Analysis of Various Design Architectures for High Safety-significant Safety-related Digital Instrumentation and Control Systems of Nuclear Power Plants during Accident Scenarios

This report documents the plus-up activities performed by Idaho National Laboratory (INL) during Fiscal Year (FY) 2022 for the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk Informed Systems Analysis (RISA) Pathway, digital instrumentation and control (DI&C) risk assessment project. In FY 2019, the RISA Pathway initiated a project to develop a risk assessment strategy for delivering a strong technical basis to support effective, licensable, and secure DI&C technologies for digital upgrades/designs. An integrated risk assessment technology for the DI&C systems was proposed for this strategy, which aims to (1) provide a best-estimate, risk-informed capability to quantitatively and accurately estimate the safety margin obtained from plant modernization, especially for the high safety-significant safety-related (HSSSR) DI&C systems, (2) support and supplement existing advanced risk-informed DI&C design guides by providing quantitative risk information and evidence, (3) offer a capability of design architecture evaluation of various DI&C systems to support system design decisions and diversity and redundancy applications, (4) assure the long-term safety and reliability of HSSSR DI&C systems, and (5) reduce uncertainty in costs and support integration of DI&C systems in the plant. To achieve these technical goals and deal with the expensive licensing justifications from regulatory insights, the LWRS-developed framework instructs nuclear vendors and utilities on how to effectively lower the costs associated with digital compliance and speed industry advances by: (1) defining an integrated risk-informed analysis process for DI&C upgrade, including hazard analysis, reliability analysis, and consequence analysis, (2) applying systematic and risk-informed tools to address common cause failures (CCFs) and quantify corresponding failure probabilities for DI&C technologies, particularly software CCFs, (3) evaluating the impact of digital failures at the component level, system level, and plant level, and (4) providing insights and suggestions on designs to manage the risks, thus to support the development, licensing, and deployment of advanced DI&C technologies on nuclear power plant (NPPs). Adding diversity within system or components is the main means to eliminate and mitigate CCFs, but diversity also increases plant complexity and errors and may not address all sources of systematic failures. How to optimize the diversity and redundancy applications for the safety-critical DI&C systems remains a challenge. To deal with the technical issues in addressing potential software CCFs in HSSSR DI&C systems of NPPs and supporting relevant design optimization, the framework provides: ? An integrated best-estimate, risk-informed capability to address new technical digital issues quantitatively, accurately, and efficiently in plan modernization progress, such as software CCFs in HSSSR DI&C systems of NPPs ? A common and a modularized platform for DI&C designers, software developers, cybersecurity analysts, and plant engineers to efficiently predict and prevent risk in the early design stage of DI&C systems ? Technical bases and risk-informed insights to assist U.S. Nuclear Regulatory Commission (NRC) and industry to address and fulfill the risk-informed alternatives for evaluation of CCFs in HSSSR DI&C systems of NPPs ? An integrated risk-informed tool that offers a capability of design architecture evaluation of various DI&C systems to support system design decisions in diversity and redundancy applications. The plus-up research and development efforts of this project in FY 2022 are focused on methodology improvement of software CCF modeling and estimation, prevention analysis, importance analysis and risk analysis of various design architectures of HSSSR DI&C systems. This work greatly enhances the capability of the LWRS-developed framework for the risk assessment and design optimization of safety-critical DI&C systems. It should be noted that all the analyses are performed for the demonstration of the LWRS-developed framework, not for the evaluation of relevant systems. Results are obtained based on very limited design information and testing data.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

An Integrated Framework for Risk Assessment of High Safety Significant Safety-related Digital Instrumentation and Control Systems in Nuclear Power Plants: Methodology and Demonstration

This report documents the activities performed by Idaho National Laboratory (INL) during Fiscal Year (FY) 2022 for the U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program, Risk Informed Systems Analysis (RISA) Pathway, digital instrumentation and control (DI&C) risk assessment project. In FY 2019, the RISA Pathway initiated a project to develop a risk assessment strategy for delivering a technical basis to support effective and secure DI&C technologies for digital upgrades/designs. A framework was proposed for this strategy, which aims to (1) provide a best-estimate, risk-informed capability to quantitatively and accurately estimate the risk impact of plant modernization, considering the introduction of high safety-significant safety-related (HSSSR) DI&C systems, (2) support and supplement existing risk-informed DI&C design guides by providing quantitative risk information and evidence, (3) offer a capability of design architecture evaluation of various DI&C systems, (4) assure the long-term safety and reliability of HSSSR DI&C systems, and (5) reduce uncertainty in costs and support integration of DI&C systems in the plant. To achieve these technical goals, the framework provides a means to address relevant technical issues by: (1) defining a risk-informed analysis process for DI&C upgrade, that integrates hazard analysis, reliability analysis, and consequence analysis, (2) applying risk-informed tools to address common cause failures (CCFs) and quantify corresponding failure probabilities for DI&C technologies, particularly software CCFs, (3) evaluating the impact of digital failures at the component level, system level, and plant level, and (4) providing insights and suggestions on designs to manage the risks, thus to support the development and deployment of advanced DI&C technologies on nuclear power plant (NPPs).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Software Dependability for Digital Instrumentation and Control Systems in Nuclear Power Plants - Lessons Learnt from a DOE Sponsored Project

This paper discusses the lessons learnt from a sponsored study undertaken under the sponsorship of the Department of Energy by a team of scientists at the Ohio State University in collaboration with Technology Resources and experts around the world. This paper makes recommendations for future research and studies based on these lessons learnt. The purpose of the research was to provide a software dependability quantification framework that would allow its evaluation along the software development life cycle for safety critical instrumentation and control systems (I and C) such as those found in nuclear power plants. A typical example of safety critical system is the reactor protection system. The research funded by the Department of Energy built on preliminary research carried out for the US NRC that had focused solely on software reliability quantification. Quantification of software dependability is important for safety critical I and C because the regulatory framework in use in the United States for nuclear power plants is risk informed. Quantification of risk requires quantification of all contributors to risk and those include instrumentation and control systems. The purpose of the study was to help developers of software-based I and C with an early evaluation of the potential risk an application would represent at different stages of the software development life cycle by building gates at the end of each life cycle phase. Life cycle phases of software development comprehend phases such as requirements, design, coding and test. The gates' function would be to stop development from moving forward if the assessed dependability objective for that phase were not met. Gates thresholds were to be defined based on measurements collected during the life cycle phase under evaluation and phases that preceded it and an understanding of the mechanisms that would lead to threatening the dependability objectives. The information contained in the gates would in addition help steer the development towards areas that needed improvement. The following are four high level conclusions to our study. First the study demonstrates a lack of consensus on mechanisms that will threaten the dependability attributes. This lack of consensus should be resolved. The lack of consensus appears for all measures to the exception of reliability. Second, measures should be developed for software availability and safety. Third, for all dependability attributes (with the exception of reliability) consistency of the measurement threads should evaluated and measurement-based development gates derived. Finally our ability to perform such type of research depends in a significant manner on access to vendor and/or utility data. The licensing environment and the first-of-a-kind nature of current software-based applications for nuclear I and C make it difficult to collect the relevant data. An approach which would either increase access to data by providing 'immunity' to the vendor or utility or the development of a software-based I and C for the sole purpose of data collection for research should be investigated. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2014

Fukushima Daiichi—A Case Study for BWR Instrumentation and Control Systems Performance During a Severe Accident, Rev 1

This report provides an assessment of the performance of various instrumentation and control (I&C) and monitoring systems, with a focus on the first few days of the Fukushima Daiichi accident starting on March 11, 2011. Similar to what occurred at the Three Mile Island Unit 2 (TMI-2) accident in 1979, operator responses were challenged when information on key parameters was lost, erroneous, or misleading. While numerous improvements in instrumentation capabilities have been made since the TMI-2 event, the accident at Fukushima Daiichi suggests the potential for further improvement. The objective of this research is to review the performance of the Fukushima Daiichi Units 1–3 instrumentation systems that could have helped ensure that plant operators and emergency responders had relevant, accurate, and timely knowledge of plant conditions as they managed the accident response. As was the case with TMI-2, forensic analyses of plant instrumentation failures will be a long-term effort. As an introduction to this long-term research, this report focuses on the information available in the control room during the accident and illustrates the challenges to operators in their understanding of plant conditions during the progression of the accident. I&C systems monitor the success or failure of achievement of light water reactor safety principles in which the protection of the reactor fuel, reactor coolant system, and reactor containment is accomplished. This report (1) provides an overview of the safety principles; (2) notes parameters associated with the boiling water reactor (BWR)/3 and BWR/4 protection, actuation, and monitoring systems as they relate to the achievement of these principles; (3) presents data from past research on predicted instrumentation performance during severe accidents, and (4) assesses performance of the key I&C and monitoring systems associated with BWR/3 and BWR/4 reactors like those at Fukushima Daiichi. Lessons learned or potential improvements associated with the performance of these systems in the context of severe accidents are delineated. As background information, an overview of key BWR/3- and BWR/4-Mark I reactor designs (BWR/3 and BWR/4 with a Mark I containment) reactor protection, engineered safety features actuation, and accident monitoring systems is provided. A brief description of plant safety systems used to prevent or mitigate reactor accidents is also provided. This background information provides context for the brief discussion of previous research that has been conducted on BWR severe accidents and for the discussion of the accident sequences, which led to core damage at Fukushima Daiichi Units 1–3. The progression of the accident sequences at Fukushima Daiichi Units 1–3 occurred very much as predicted in prior severe accident research conducted in the United States and internationally, given the complete station blackout that occurred when the tsunami generated by one of the most powerful earthquakes ever recorded rolled onshore.

Accident Tolerant Equipment↗
NASA NTRS2005

Thermal Assessment of Swift BAT Instrument Thermal Control System in Flight

THE BAT is the primary instrument on the Swift spacecraft. The Swift mission is part of the National Aeronautics and Space Administration (NASA) Medium-Size Explorer (MIDEX) Program, and is managed by Goddard Space Flight Center (GSFC). It is designed to detect gamma ray burst over a broad region of the sky in a low Earth orbit of 600-km altitude and quickly align the telescopes on the spacecraft to the gamma ray source. It was successfully launched into orbit on November 20, 2004. The Swift mission is a first of its kind of multi-wavelength transient observatory for gamma ray burst astronomy. Its mission life is 2 years. The inclination is 22 deg maximum. The spacecraft bus voltage to the instruments is in the 24 V to 35 V range. The instruments will be turned off when the voltage is below 27 V. The BAT is mounted to the optical bench through five titanium flexures. The BAT has been developed at GSFC. Its telescope assembly consists of 256 Detector Modules (DMs) in the Detector Array. There are 16 Detector Array Blocks. Each Block holds 16 DMs, 3 Block Voltage Regulator (BVR) units and 3 Block Command & Data Handling (BCDH) units. The power dissipation of each Block has been measured to be 13 W. Therefore the total power dissipation of the 16 Blocks is 208 W. The DAP is 1.3 m (4.3 ft) x 1 m (3.3 ft), accommodates all the 16 Blocks. It also provides the mounting surface and the positional stability for the Blocks. The DMs are located at the top (+X side) of the DAP and is enclosed by graded-Z shields on the sides and a coded mask at the top. The BVRs and BCDHs are located at the bottom (-X side) of the DAP. Eight Blocks are located at the front (-Z side or radiator side) of the DAP, and eight are located at the rear (+Z side) of the DAP. The DMs and top of DAP are insulated with a 7-layer multi-layer insulation (MLI). There is a 5.08 cm (2 in) x 5.08 cm (2 in) MLI cutout over each Block heater controller so that heat radiates from the heater controller to the mask. The exterior of the mask, graded-Z shields and bottom of DAP is insulated with a 15-layer MLI.

Choi, Michael K.↗
DOE OSTI2023

Analyzing Hardware and Software Common Cause Failures in Digital Instrumentation and Control Systems using Dual Error Propagation Method

This paper develops a methodology for quantifying software common cause failures (CCFs) in digital instrumentation and control (I&C) systems of nuclear power plants. To support the transition of analog I&C systems to digital in nuclear power plants, probabilistic risk assessment (PRA) techniques are used. The hardware components of the I&C systems have reliability databases that can be used in the PRA studies. However, the failure data for redundant software components of the systems is sparse. Failure of components constitutes a CCF, wherein two or more components or systems fail due to a single shared cause and coupling mechanism. This paper proposes a quantification approach that can simultaneously model hardware and software components, incorporate the CCFs of software systems in the models, and bridge the gap between the failure quantification of models and the development of CCF parametric databases. We demonstrate the dual error propagation method (DEPM) by developing I&C systems failure models for a representative digital reactor trip system. The DEPM models are built to simulate the control and data flows within the systems and can accommodate failure states. By expanding DEPM to software CCFs, we generated alpha factor parameter estimates for each of the modeled error propagation mechanisms.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Dependability Assessment of Software for Safety Instrumentation and Control Systems

In its 2013 recommendations to the International Atomic Energy Agency (IAEA), the Technical Working Group on Nuclear Power Plant Instrumentation and Control (TWG-NPPIC) identified assessment and evaluation of software dependability as a topic of interest for IAEA Member States. To implement the recommendation of the TWG-NPPIC, the IAEA Scientific Secretary for the TWG initiated a consultancy activity in 2014 to develop a technical report describing a dependability assessment framework suitable for safety instrumentation and control (I and C) systems in nuclear power plants (NPPs). The development of the software dependability assessment report has involved three consultancy meetings at IAEA headquarters in Vienna, Austria. The conduct of these meetings allowed consultation among subject matter experts to generate consensus technical information that serves as the content of the report. The technical contributors consist of international experts from academia, industry, regulatory bodies and research organizations. In September 2014, a technical meeting was held in Daejeon, Republic of Korea, to present interim findings of the consultancy effort and to solicit input from representatives of IAEA Member States. The information and technical comments contributed during the technical meeting were addressed in subsequent development of the report. This paper gives an overview of key technical elements of the IAEA report. The IAEA report presents a framework by which a dependability assessment of safety software can be performed. The specified guiding principles provide the basis for establishing an assessment strategy and implementing the assessment process. The assessment framework provides a systematic, coherent structure within which to organize technical and process evidence to satisfy arguments in support of claims about software dependability properties. During the consultancy activity, the participating subject matter experts focused on high level principles, a structured approach, and identification of suitable methods and techniques. However, it was recognized that what remains is to concretize the framework through specific examples and usage experience. Thus, following publication of the report, it is recommended that case studies be performed to illustrate and more precisely define specific phases and elements of the assessment framework. For example, current national assessment approaches could be mapped into the assessment framework to serve as case studies and help to identify any aspects of the assessment framework that needs further definition or adjustment. Additionally, a challenge problem could be defined to engage stakeholders to apply the assessment framework with a common focus. This could help to better illustrate strategic options and approaches to the assessment phases. It also could serve to suggest other considerations and potential sources of evidence not fully covered in the report. Based on findings from case studies, challenge problems and country experience, it should be possible to develop generalized templates to support application of the assessment framework. These findings can be collected to serve as the basis for a User's Guide on dependability assessment of safety software. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

Development of functional requirements specification for digital instrumentation and control systems upgrades used at nuclear power plants (NPPs) - 396

Recent market conditions have driven the nuclear industry to become more innovative at reducing cost in order to remain competitive. Instrumentation and Control (I and C) systems used by many NPPs are based on outdated 1960's and 70's technology. Many of the components used in these digital systems are no longer supported by Original Equipment Manufacturers (OEMs) and are rapidly becoming obsolete. Coping with the high cost of replacing I and C equipment due to obsolescence adds to this economic challenge. While other industries have migrated from analog to modern digital I and C technology, implementation of digital I and C controls within the nuclear industry has been very slow. This is primarily due to the high capital cost related to implementation of digital I and C upgrades, which is compounded by regulatory uncertainties that present significant licensing risks. The Nuclear Regulatory Commission (NRC), Nuclear Energy Institute (NEI) and the industry's Digital I and C Working Group are working to address many of the regulatory barriers that adversely impacts implementation of digital upgrades at NPPs. The NRC has developed an 'Integrated Action Plan to Modernize Digital Instrumentation and Controls Regulatory Infrastructure' (ML16097A182), Reference 1. Once fully implemented, this plan is expected to help reduce regulatory uncertainties that affects project cost. Excluding regulatory uncertainty, most OEM's will concur that one of the most important factors in reducing project cost is assuring that all requirements are well defined and complete prior to starting the project. A detailed and high quality Functional Requirements Specification (FRS) is critical in documenting these requirements. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Installation of the Dark Energy Spectroscopic Instrument at the Mayall 4-meter telescope

The Dark Energy Spectroscopic Instrument (DESI) is a Stage IV ground-based dark energy experiment that will measure the expansion history of the Universe using the Baryon Acoustic Oscillation technique. The spectra of 35 million galaxies and quasars over 14000 square degrees will be measured during the life of the experiment. We describe the installation of the major elements of the instrument at the Mayall 4m telescope, completed in late 2019. The previous prime focus corrector, spider vanes, and upper rings were removed from the Mayall’s Serrurier truss and replaced with the newlyconstructed DESI ring, vanes, cage, hexapod, and optical corrector. The new corrector was optically aligned with the primary mirror using a laser tracker system. The DESI focal plane system was integrated to the corrector, with each of its ten 500-fiber-positioner petal segments installed using custom installation hardware and the laser tracker. Ten DESI spectrographs with 30 cryostats were installed in a newly assembled clean room in the Large Coude Room. The ten cables carrying 5000 optical fibers from the positioners in the focal plane were routed down the telescope through cable wraps at the declination and hour angle axes, and their integral slitheads were integrated with the ten spectrographs. The fiber view camera assembly was installed to the Mayall’s primary mirror cell. Servers for the instrument control system replaced existing computer equipment. The fully integrated instrument has been commissioned and is ready to start its operations phase.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRS1974

Sneak circuit analysis of instrumentation and control systems

The evolvement and implementation is presented of a systematic means for detecting the sneak prone designs of an electrical system which could result in such unplanned modes so that they can be prevented prior to occurrence. Sneak circuit analysis is shown to be a formalized approach based upon topological techniques. The trees employed are produced from manufacturing detail data processed by a computer to orderly completion. The analysis techniques applied on a variety of NASA programs are shown to be particularly applicable to industrial instrumentation and control systems.

Rankin, J. P.↗
NASA NTRS2008

Swift BAT Instrument Thermal Control System Recovery after Spacecraft Safehold in August 2007

The Swift mission Burst Alert Telescope (BAT) Detector Array thermal control system includes two propylene loop heat pipes (LHPs), eight ammonia constant conductance heat pipes (CCHPs), a radiator that has AZ-Tek's AZW-LA-II low alpha white paint, and precision heater controllers that have adjustable set points in flight. The Power Converter Box (PCB) and Image Processor Electronics (IPE) boxes (a primary and a redundant) of the BAT have Z93P white paint radiators. Swift was successfully launched into orbit on November 20, 2004. The spacecraft (S/C) was placed into a safehold mode on August 10, 2007 after an anomaly on inertial reference unit (IRU) #3. It was returned to inertial pointing on August 16 and instrument power up followed. This paper presents a thermal assessment of the BAT instrument thermal control system (TCS) shut down and recovery as a result of the S/C safehold mode. The recovery required starting up the LHPs manually.

Choi, Michael K.↗
DOE OSTI2012

Light Water Reactor Sustainability Program: Advanced Instrumentation, Information, and Control Systems Technologies Technical Program Plan

Reliable instrumentation, information, and control (II&C) systems technologies are essential to ensuring safe and efficient operation of the U.S. light water reactor (LWR) fleet. These technologies affect every aspect of nuclear power plant (NPP) and balance-of-plant operations. In 1997, the National Research Council conducted a study concerning the challenges involved in modernization of digital instrumentation and control systems in NPPs. Their findings identified the need for new II&C technology integration.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

ProtoDESI: First On-Sky Technology Demonstration for the Dark Energy Spectroscopic Instrument

The Dark Energy Spectroscopic Instrument (DESI) is under construction to measure the expansion history of the universe using the baryon acoustic oscillations technique. The spectra of 35 million galaxies and quasars over 14,000 square degrees will be measured during a 5-year survey. A new prime focus corrector for the Mayall telescope at Kitt Peak National Observatory will deliver light to 5,000 individually targeted fiber-fed robotic positioners. The fibers in turn feed ten broadband multi-object spectrographs. We describe the ProtoDESI experiment, that was installed and commissioned on the 4-m Mayall telescope from 2016 August 14 to September 30. ProtoDESI was an on-sky technology demonstration with the goal to reduce technical risks associated with aligning optical fibers with targets using robotic fiber positioners and maintaining the stability required to operate DESI. The ProtoDESI prime focus instrument, consisting of three fiber positioners, illuminated fiducials, and a guide camera, was installed behind the existing Mosaic corrector on the Mayall telescope. A fiber view camera was mounted in the Cassegrain cage of the telescope and provided feedback metrology for positioning the fibers. ProtoDESI also provided a platform for early integration of hardware with the DESI Instrument Control System that controls the subsystems, provides communication with the Telescope Control System, and collects instrument telemetry data. In conclusion, lacking a spectrograph, ProtoDESI monitored the output of the fibers using a fiber photometry camera mounted on the prime focus instrument. ProtoDESI was successful in acquiring targets with the robotically positioned fibers and demonstrated that the DESI guiding requirements can be met.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2019

Achieving Verifiable and High Integrity Instrumentation and Control Systems through Complexity Awareness and Constrained Design. Final Report

While most process control industries have been able to embrace digital Instrumentation and Control (I&C) technology to improve performance, reliability, maintainability, and efficiency of production, the nuclear industry has been relatively slow to adopt digital I&C for safety critical plant functions. There are a variety of reasons for this, but in general the nuclear industry takes a conservative stance that new technologies must mature before adoption. A deeper underlying reason for not fully realizing digital I&C and automation in nuclear power is associated with the complexity that can manifest in software-based systems that is different in nature from the complexity of their analog predecessors. Accordingly, this research pursued the development of innovative architectures and methods to “constrain complexity” to enhance verifiability of digital I&C devices. These constraints can positively impact the reduction of Software Common Cause Failures (SCCF) and other systematic errors. The architectural concept developed in this research effort is called SymPLe. SymPLe is specifically aimed at safety critical functions within Nuclear Power Plants and other critical infrastructure where constraining device complexity to support verifiability and safety case arguments is important. The SymPLe architecture is designed to promote “accessible verification” between OEM/utility engineers and the regulatory agencies who review and license nuclear power plant I&C systems. The major contributions of this work to the nuclear industry are: Demonstration of SymPLe as a viable approach to “complexity aware” I&C systems and Embedded Digital Devices.Comprehensive use of model-based design and engineering to produce verifiable I&C systems for the nuclear industry with clear findings on the usability of model-based engineering. Employ comprehensive use of model-based engineering, testing, and verification in an IEC 61508 workflow in support of design assurance. Gathered evidence of “difficult to find” design flaws found using model-based design and testing methods. Detailed the nature of these SCCFs including where they were introduced, the triggers, and how they were found. Gathered evidence on the synergistic use of practical formal methods and testing and how these methods can be used to significantly increase assurance of reduction and avoidance of design flaws.

Testing↗