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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
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

Multiple aspects maintenance ontology-based intelligent maintenance optimization framework for safety-critical systems

Abstract Maintenance optimization is a process for improving the efficiency of maintenance strategies and activities, considering various aspects of the target system and components, such as the probabilities of system failures and the cost of repair and replacement of a failed component. The improvement of maintenance optimization algorithms generally requires information from various data sources. For example, it may require the system risk information derived from risk analysis tools or the residual lifetime of a component from fault prognosis tools. The requirements of data acquisition (DAQ) and aggregation pose new challenges for maintenance management systems (MMSs) that implement and use these maintenance optimization algorithms. This paper proposes a multiple aspects maintenance ontology-based framework to facilitate DAQ from MMSs, online monitoring systems, fault detection and discrimination tools, risk assessment tools, decision-making tools, and component identification tools, and accelerate the implementation and verification of contemporary maintenance optimization models and algorithms. The proposed framework consists of a multi-aspect maintenance ontology with critical information for maintenance optimization and application interfaces for collecting information from various data sources, such as fault prognosis tools, online monitoring tools, risk assessment tools, and decision-making algorithms. In addition, this paper proposes a heuristic method for integrating concepts and properties from other existing ontologies into the proposed framework when the existing ontology is not fully compatible with the ontology under construction. Finally, the paper verifies the proposed ontology framework using a feedwater system designed for nuclear power plants with valves and filters as the components under maintenance.

Diao, Xiaoxu (ORCID:0000000346726352)↗
DOE OSTI2017

Risk assessment of safety-critical data communication in digital safety feature control system - 205

As one of the safety-critical systems in Advanced Power Reactor-1400 (APR-1400) nuclear power plant (NPP), the Engineered Safety Feature-Component Control System (ESF-CCS) employs safety data link and network communication for transmitting safety component actuation to effectively facilitate various safety-critical field controllers. However, data communication failure risk in the ESF-CCS has yet to be fully quantified, the probability that a safety-critical system in digitalized NPP becomes unsafe due to a network failure must be evaluated to quantify the risk of digital I and C system. In this study, a fault tree model was developed to assess the data link and data network failure-induced unavailability of a safety-critical system function used to generate an automated control signal for actuating design basis accident mitigation equipment. Especially, the risk of data communication failure in a digital safety feature control system was analyzed in consideration of interconnection between safety-critical controllers and the fault-tolerant algorithm implemented in the network configuration. Based on the developed fault tree model, case studies were performed to quantitatively assess the unavailability of ESF-CCS signal generation due to safety data link and network failure and their risk effect on safety signal generation unavailability. This study is expected to provide risk information on the safety-critical data communication in a digitalized NPP instrumentation and control system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

R-matrix evaluation of the 15 N Compound System for Criticality Safety Analyses [Abstract]

The evaluation of the 15 N compound system was selected by the Nuclear Criticality Safety Program (NCSP) due to the importance of nitrogen in criticality benchmarks and in actinide chemistry encountered during fuel reprocessing. For improved accuracy in transport calculations, it is recommended to obtain and distribute resonance parameters to describe n + 14 N reactions.

AZURE↗
DOE OSTI2019

Bus based timed input output module

Described herein are features related to a bus based timed input output module (TIO) for use in control systems of physical systems, particularly where the physical systems are safety critical systems. When the TIO is powered on, one or more operations are performed by the TIO, wherein the one or more operations, when performed by the TIO: 1) cause the TIO to be non-functional, thereby preventing the TIO from improperly or accidentally activating a Component in communication with the TIO; 2) clears the TIO of previous operations (if any); and 3) facilitates compatibility of the TIO with an arbitrary Component.

42 ENGINEERING↗
DOE OSTI2022

An ontology-based fault generation and fault propagation analysis approach for safety-critical computer systems at the design stage

Abstract Fault propagation analysis is a process used to determine the consequences of faults residing in a computer system. A typical computer system consists of diverse components (e.g., electronic and software components), thus, the faults contained in these components tend to possess diverse characteristics. How to describe and model such diverse faults, and further determine fault propagation through different components are challenging problems to be addressed in the fault propagation analysis. This paper proposes an ontology-based approach, which is an integrated method allowing for the generation, injection, and propagation through inference of diverse faults at an early stage of the design of a computer system. The results generated by the proposed framework can verify system robustness and identify safety and reliability risks with limited design level information. In this paper, we propose an ontological framework and its application to analyze an example safety-critical computer system. The analysis result shows that the proposed framework is capable of inferring fault propagation paths through software and hardware components and is effective in predicting the impact of faults.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

Final Technical Report on Quantifying Dependability Attributes of Software Based Safety Critical Instrumentation and Control Systems in Nuclear Power Plants

With the current transition from analog to digital instrumentation and control systems in nuclear power plants, the number and variety of software-based systems have significantly increased. The sophisticated nature and increasing complexity of software raises trust in these systems as a significant challenge. The trust placed in a software system is typically termed software dependability. Software dependability analysis faces uncommon challenges since software systems’ characteristics differ from those of hardware systems. The lack of systematic science-based methods for quantifying the dependability attributes in software-based instrumentation as well as control systems in safety critical applications has proved itself to be a significant inhibitor to the expanded use of modern digital technology in the nuclear industry. Dependability refers to the ability of a system to deliver a service that can be trusted. Dependability is commonly considered as a general concept that encompasses different attributes, e.g., reliability, safety, security, availability and maintainability. Dependability research has progressed significantly over the last few decades. For example, various assessment models and/or design approaches have been proposed for software reliability, software availability and software maintainability. Advances have also been made to integrate multiple dependability attributes, e.g., integrating security with other dependability attributes, measuring availability and maintainability, modeling reliability and availability, quantifying reliability and security, exploring the dependencies between security and safety and developing integrated analysis models. However, there is still a lack of understanding of the dependencies between various dependability attributes as a whole and of how such dependencies are formed. To address the need for quantification and give a more objective basis to the review process -- therefore reducing regulatory uncertainty -- measures and methods are needed to assess dependability attributes early on, as well as throughout the life-cycle process of software development. In this research, extensive expert opinion elicitation is used to identify the measures and methods for assessing software dependability. Semi-structured questionnaires were designed to elicit expert knowledge. A new notation system, Causal Mechanism Graphing, was developed to extract and represent such knowledge. The Causal Mechanism Graphs were merged, thus, obtaining the consensus knowledge shared by the domain experts. In this report, we focus on how software contributes to dependability. However, software dependability is not discussed separately from the context of systems or socio-technical systems. Specifically, this report focuses on software dependability, reliability, safety, security, availability, and maintainability. Our research was conducted in the sequence of stages found below. Each stage is further examined in its corresponding chapter. Stage 1 (Chapter 2): Elicitation of causal maps describing the dependencies between dependability attributes. These causal maps were constructed using expert opinion elicitation. This chapter describes the expert opinion elicitation process, the questionnaire design, the causal map construction method and the causal maps obtained. Stage 2 (Chapter 3): Elicitation of the causal map describing the occurrence of the event of interest for each dependability attribute. The causal mechanisms for the “event of interest” were extracted for each of the software dependability attributes. The “event of interest” for a dependability attribute is generally considered to be the “attribute failure”, e.g. security failure. The extraction was based on the analysis of expert elicitation results obtained in Stage 1. Stage 3 (Chapter 4): Identification of relevant measurements. Measures for the “events of interest” and their causal mechanisms were obtained from expert opinion elicitation for each of the software dependability attributes. The measures extracted are presented in this chapter. Stage 4 (Chapter 5): Assessment of the coverage of the causal maps via measures. Coverage was assessed to determine whether the measures obtained were sufficient to quantify software dependability, and what measures are further required. Stage 5 (Chapter 6): Identification of “missing” measures and measurement approaches for concepts not covered. New measures, for concepts that had not been covered sufficiently as determined in Stage 4, were identified using supplementary expert opinion elicitation as well as literature reviews. Stage 6 (Chapter 7): Building of a detailed quantification model based on the causal maps and measurements obtained. Ability to derive such a quantification model shows that the causal models and measurements derived from the previous stages (Stage 1 to Stage 5) can form the technical basis for developing dependability quantification models. Scope restrictions have led us to prioritize this demonstration effort. The demonstration was focused on a critical system, i.e. the reactor protection system. For this system, a ranking of the software dependability attributes by nuclear stakeholders was developed. As expected for this application, the stakeholder ranking identified safety as the most critical attribute to be quantified. A safety quantification model limited to the requirements phase of development was built. Two case studies were conducted for verification. A preliminary control gate for software safety for the requirements stage was proposed and applied to the first case study. The control gate allows a cost effective selection of the duration of the requirements phase.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Development of a generalized perturbation theory method for sensitivity analysis using continuous-energy Monte Carlo methods

The sensitivity and uncertainty analysis tools of the ORNL SCALE nuclear modeling and simulation code system that have been developed over the last decade have proven indispensable for numerous application and design studies for nuclear criticality safety and reactor physics. SCALE contains tools for analyzing the uncertainty in the eigenvalue of critical systems, but cannot quantify uncertainty in important neutronic parameters such as multigroup cross sections, fuel fission rates, activation rates, and neutron fluence rates with realistic three-dimensional Monte Carlo simulations. A more complete understanding of the sources of uncertainty in these design-limiting parameters could lead to improvements in process optimization, reactor safety, and help inform regulators when setting operational safety margins. A novel approach for calculating eigenvalue sensitivity coefficients, known as the CLUTCH method, was recently explored as academic research and has been found to accurately and rapidly calculate sensitivity coefficients in criticality safety applications. The work presented here describes a new method, known as the GEAR-MC method, which extends the CLUTCH theory for calculating eigenvalue sensitivity coefficients to enable sensitivity coefficient calculations and uncertainty analysis for a generalized set of neutronic responses using high-fidelity continuous-energy Monte Carlo calculations. Here, several criticality safety systems were examined to demonstrate proof of principle for the GEAR-MC method, and GEAR-MC was seen to produce response sensitivity coefficients that agreed well with reference direct perturbation sensitivity coefficients.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Security-informed safety: integrating security within the safety demonstration of a smart device - 157

Safety and security engineering have, over the years, developed their own regulations, standards, cultures, and practices. However, there's a growing realisation that security is closely connected to safety. Safety must be security-informed: if a safety-critical system isn't secure, it isn't safe. A safety demonstration is incomplete and unconvincing unless it considers security. In our work for government and industry, we have used the Claims, Arguments, Evidence (CAE) framework to analyse the impact of security on a safety justification or safety case and identified the significant changes needed to address security explicitly. This will impact the design and implementation process as well as the assurance and V and V approach. In this paper we discuss the impact of integrating security when developing a safety demonstration of a smart device. A smart device is an instrument, device or component that contains a microprocessor (and therefore contains both hardware and software) and is programmed to provide specialised capabilities, often measuring or controlling a process variable. Examples of smart devices include radiation monitors, relays, turbine governors, uninterruptible power supplies and heating ventilation, and air conditioning controllers. (authors)

42 ENGINEERING↗
DOE OSTI2017

SymPLe 1131: A novel architectural solution for the realization of verifiable digital I and C systems and embedded digital devices - 387

One of the main reasons for the increasing popularity of FPGAs in embedded systems and digital I and C systems is that they provide many of the advantages from both hardware and software components. FPGAs provide configurability approaching that of SW, while at the same time delivering deterministic behavior of pure hardware solutions. However, with regard to Software Common Cause Failures (CCF), the use of FPGAs in digital I and C platforms should not be seen as plug-in remedy for reduction or elimination of CCF. To address CCF, we argue that it is not enough to manage the complexity of an I and C device or system, but we must also have the ability to constrain complexity (as needed) and rigorously reason about execution and device functionality. This paper presents the work done regarding an FPGA overlay architecture called SymPLe. SymPLe is designed to reduce the gap between software and hardware approaches for digital I and C systems and increases deterministic behavior and verifiability of the application, architecture and safety-critical system while reasoning about CCF. (authors)

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Non-dimensional performance and safety parameters for heat pipes

The use of heat pipes in safety-critical systems such as nuclear microreactors dictates the development of generalized, practical, scalable performance and safety parameters. Traditional dimensional metrics, while informative, lack the universality required for comparative analysis across varying designs and operating regimes. Here, this work introduces a comprehensive set of non-dimensional parameters to characterize heat pipe performance and safety, including capillary performance, effective thermal conductivity, response time, exergetic efficiency, allowable temperature gradients, allowable rate of temperature change, priming coefficients, and factor of safety. A reference heat pipe design representative of microreactor applications was analyzed via the developed parameters using both traditional analytical models and Sockeye simulations under transient and steady-state conditions. Sodium, potassium, and water were evaluated as working fluids to demonstrate the applicability of the framework across a broad temperature range. The proposed non-dimensional parameters effectively captured key thermal-hydraulic behaviors and safety concerns, as was demonstrated via Sockeye simulations. This framework supports the development of design optimization strategies, operational protocols, and safety assurance practices for advanced reactor systems and other high-reliability applications.

42 - ENGINEERING↗
DOE OSTI2017

NRC technical basis for evaluation of its position on protection against common cause failure in digital systems used in nuclear power plants - 378

Digital technology has advantages over analog systems, including automated monitoring and alerts for standby safety functions, and predictive algorithms to maintain critical safety systems. Additionally, digital technology generally has higher reliability and can be designed to reduce single point vulnerabilities. For these reasons many nuclear plants have applied digital technology to safety and non-safety related applications, including reactor protection system, feedwater and turbine controls, etc. with a corresponding significant improvement in trip reduction. Nonetheless, digital instrumentation and control (I and C) systems also present potential new vulnerabilities that need to be assessed, including potential failures due to increased complexity of digital systems, the introduction of unique failure modes due to software (including software common cause failure (CCF)), and limited operating history of digital systems in nuclear safety related applications compared to analog systems. The fact that software is intangible means that common methods, such as analysis or testing, used for detecting CCF may not be effective when applied to software. Consequently, digital technology is perceived to pose a potential risk from the introduction of undetected systematic faults that could result in CCF. Despite the I and C system upgrades and modifications performed to date, the U.S. Nuclear Regulatory Commission (NRC) and industry stakeholders have identified the need to modernize the regulatory infrastructure to efficiently address risks associated with the use of digital technology for nuclear safety applications and address regulatory uncertainties. The NRC's current position on CCF is guided by the staff requirements memorandum (SRM) on SECY 93-087. The SRM provides specific acceptance criteria for the evaluation of CCF, which the staff implemented in the Branch Technical Position (BTP) 7-19. However, industry stakeholders have proposed using methods to characterize the likelihood of software CCF and eliminate it from further consideration in a defense-in-depth and diversity analysis. The NRC's current position does not consider these alternatives, and thus corresponding acceptance criteria is not currently available. The work discussed in this paper assesses the underlying technical basis associated with CCF, provides technical support for updating the NRC position and considers proposed methods for addressing potential CCF in digital systems while enhancing efficiency, clarity, and confidence. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

A Novel Methodology For Generating Thermal Scattering Cross Sections and Uncertainties

With the rise in interest of GEN-IV reactor systems, newer, and more accurate thermal scattering nuclear data are needed. Both GEN-IV reactors and current light water reactors applying for license extensions require high fidelity cross sections and uncertainties to better thermal margins and thereby maximizing energy production and improving economics and safety. Thermal moderator data play a key role in both reactor physics and nuclear criticality safety analyses. Due to the lack of uncertainties data for thermal scattering material, there is no way to quantify the effects of thermal scattering uncertainties in quantities of interest (e.g., k eff ) in reactor or criticality safety systems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Enhancing Operational Safety via Agentic Dialogue Hazard Identification Analysis

Operational safety in high-stakes domains such as industrial process control, autonomous, and safety-critical systems demand reliable hazard identification. While large language models (LLMs) have shown promise in automating safety analysis tasks, single-turn, monolithic inference is brittle: it lacks the self-correction, deliberation, and contextual refinement that safety engineers apply iteratively. In this paper, we introduce HAZDIAL, a framework that investigates whether structured agentic dialogue (multi-agent, multi-turn interactions) improves the quality of NLP-based hazard identification over single-pass baselines. We systematically compare two dialogue modalities: adversarial debate and constructive discussion, and propose an genetic algorithm-based agentic interaction optimization. We evaluate all configurations against a curated golden dataset using standard classification metrics (accuracy, precision, recall, F1) and a novel dialogue metrics. This work advances the intersection of dialogue systems, multi-agent reasoning, and AI safety, providing empirical evidence for dialogue-driven hazard analysis.

Das, Sanjay [ORNL] (ORCID:0009000542591915)↗
DOE OSTI2016

Modeling the Resilience of Nuclear Power Plant for Unexpected Reactor Trip

Resilience can be defined as the intrinsic ability of a system to adjust to its functioning prior to, during, or following changes and disturbances, so that it can sustain required operations under both expected and unexpected conditions. Resilience engineering is a relatively new paradigm for safety management that focuses on how to cope with complexity under pressure or disturbance to achieve success. A resilient system combines inherent human abilities with engineered safety features and organizational features in a controlled and flexible manner both in expected and in unexpected situations. The concept of resilience has been suggested to complement conventional safety analyses. Conventional safety analysis methods such as PSA have several limitations: 1) they primarily focus on technical dimension, 2) the analysis are linear and sequential, 3) they are dominated by static models, 4) they do not take a systemic view into account, and 5) they focus primarily on why accidents happen and not how success is achieved. The concept of resilience provides a more integrated view for the safety analysis in safety-critical systems, e.g., nuclear power plants. The concept of resilience in the organization of NPPs has been highlighted since the Fukushima Daiichi Nuclear Power Plant accident. An IAEA report addresses that a resilient organization is one that quickly realizes deviation from normal operations and has the ability to make even the toughest and least popular decisions and to manage the margins in which it can maneuver. While traditional strategy to safety in NPPs is to identify what could go or has gone wrong, the concept of resilience is focused on what the organization does well and what it does to maintain its successful operation and to capitalize on those processes in the event of an unexpected situation. This study attempts to model the resilience of NPP dynamically, based on the operating experiences in Korean NPPs. First, this study revised the characterization of NPP. Then, a survey on the literature was conducted to identify the influences between the elements in the model of resilience in NPPs. These works provide a qualitative model for the resilience in NPPs. Then, this study reviewed 168 licensee events reports submitted to the Korean regulatory body since 2004 and quantified the relation between the attributes in the model by performing a Pearson correlation analysis. Finally, a few simulations using the System Dynamics were conducted to discuss which factors are most influencing the resilience. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗