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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 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
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↗
DOE OSTI2017

Investigation of instrumentation containing an embedded digital device - 257

The U.S. nuclear power industry primarily relies on mature, well-proven analog technology as the basis for most instrumentation and control (I and C) equipment employed in its nuclear power plants (NPPs). However, the market for this legacy instrumentation is very small and availability of such equipment for spare parts or replacement is increasingly limited. Frequently, modern I and C equipment contains embedded digital devices (EDDs), such as microprocessors and programmable logic devices. While the enhanced functional capabilities provided by the digital components offer improved performance, reliability, and maintainability, the corresponding complexity of sophisticated digital implementations also exacerbate concerns about an increased potential for undetected systematic faults, which can lead to common-cause failure (CCF). As part of a research project sponsored by the U.S. Department of Energy, an investigation has been conducted of commercially available instrumentation marketed for nuclear or industrial application. The objective was to identify the usage of EDDs and to determine the types of functional roles allocated to the devices. Examples of the equipment surveyed include smart sensors, sequencers for emergency diesel generators, pumps, valve actuators, motor control centers, breakers, priority logic modules, and uninterruptible power sources. Where sufficient detail was found, an evaluation was performed to ascertain the nature and role of the digital technology implemented in the instrumentation. The information gained through this investigation serves as technical input for defining a classification approach for EDDs based on the functional impact of the device. For example, the role of an EDD can involve any of the following functions: performance monitoring and diagnostics, information extraction and communication, instrument condition monitoring and protection, or execution of core instrument functions. Classifying digital devices according to their role in the operation of the instrumentation in which they are embedded can enable determination of the potential functional impact of failure of that device and contribute to determination of the potential safety significance of a CCF event disabling all instances of that device. This paper will report on the findings of the investigation of instrumentation with EDDs and indicate considerations for the emerging classification approach. The higher-level goal of this effort is to determine whether a graded approach to qualification testing may be possible based on classification. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Application of HFE to the design and V and V of a computerized procedure system: benefits and challenges - 234

The application of Human Factors Engineering (HFE) practices to the Human Machine Interface design and subsequent Verification and Validation (V and V) is a widely applied practice when talking about the design of complete control rooms or important control room modernizations, but not so often applied to the design of computerized operator support systems. In order to overcome the challenges faced on designing a Computerized Operation Support System (COSS) that meets the process efficiency requirements and integrates seemingly with hybrid plant operation concept, the involvement of a HFE team early in the COSS design and implementation process along with operation, Instrumentation and Control (I and C) and cybersecurity experts will guarantee the project success. This paper documents the HFE process applied during the design and evaluation of a Computerized Procedure System (CPS) and how final product was shaped to ensure that the transition from paper to computerized procedures will help in meeting Delivering the Nuclear Promise (DNP) goals, improving plant efficiency. The HFE program included the following activities: 1. A review of existing operating experience 2. An analysis of functions and tasks associated to the use and maintenance of CPS 3. A design verification of the system to meet guidelines prescribed in the NUREG-0700 (Rev.2), requirements of the ISG-05 and cybersecurity criteria (NEI 08-09, IEC-62646, EPRI 1015313, these among others). 4. An integrated system validation in a full scope simulator The application of HFE practices in the design and implementation of COSS helps not only to ensure that efficiency and safety improvement requirements are met, but also to evaluate the potential return of the investment, that such tools should bring. (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 OSTI2023

Depletion-driven thermochemistry of molten salt reactors: review, method, and analysis

Molten salt reactors (MSRs) are innovative advanced nuclear reactors that utilize nuclear fuel by dissolving it in a high-temperature liquid salt. This unique feature differentiates MSRs from other types of reactors and allows for enhanced safety and economic performance. The liquid fuel also entails several multiphysics effects that can complicate reactor design and operation. One primary effect termed here as depletion-driven thermochemistry is a driving force in altering the multiphysics behavior of the reactor. Essentially, depletion-driven thermochemistry is the effect that fuel depletion has on changing the chemical redox potential of the fuel salt over time. As the fuel is consumed, the redox potential shifts toward a more oxidizing state. Without active control, the changing chemistry due to depletion increases corrosion thereby limiting reactor component lifetimes. Additionally, the changing redox potential of the fuel salt alters the vapor pressures of chemical species dissolved in the fuel salt. Changing vapor pressures of species in the fuel salt is an important parameter to understand when off-gassing volatile species during normal reactor operation, and for source term characterization during accident scenario transients. The present work represents a fundamental step toward modeling and coupling the driving physics (i.e., neutronics and chemistry) involved in altering the redox potential in an MSR. Here, the neutronic code Griffin models the depletion of the fuel-salt system, while the chemical equilibrium code Thermochimica calculates the thermochemical state of the isotopic inventory, using the Molten Salt Thermodynamic Database - Thermochemical (MSTDB-TC). These two codes are tightly coupled to predict the impact of fuel depletion in altering the chemistry in MSR systems. Redox potential control methods are discussed and can be modeled using this multiphysics approach. The vapor pressures of chemical species that could be extracted to an off-gas system, as determined by the reactor’s thermochemical state, are examined. The neutronics-chemistry coupling developed in this work is expected to have potential application for analyzing corrosion, source term evolution, and material safeguards in MSR systems. Lastly, suggestions for areas of further improvements of the models to expand these capabilities by incorporating other coupled physics effects is provided.

Walker, Samuel A.↗
DOE OSTI2017

Severe accident: hardening of standards and qualification requirements for nuclear pressure transmitters - 319

Nuclear power stations are designed according to the nuclear safety principle: the achievement of proper operating conditions, prevention of accidents or mitigation of accident consequences, resulting in protection of workers, the public and the environment from undue radiation hazards. This principle is applicable from the design to the dismantling of the power station, including normal and accident (and post-accident) operation. To take into account the world experience (incident or accident as Fukushima), international safety authorities recommends to reinforce safety functions and qualification requirements level. Concerning the instrumentation, devices like sensors are mainly impacted by the following constraint improvements: - Radiation resistance in normal and accident conditions: - Higher total dose; - Dose Rate; - Nature of radiation; - Pressure and Temperature profile in terms of: - Magnitude of pressure and / or temperature (higher); - Duration of higher constraints (longer). All these specifications induce: - The creation of a new qualification level - Severe Accident (SA) or equivalent; - An impact on the sensor technology to improve the resistance to harsh environment. Finally, the paper will present in details the evolution of the requirements and will sum up the advantages and disadvantages of different technologies to meet the severe accident conditions. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRSDate not supplied

Kinetic Model Development of the Carbon Vapor Deposition Reactor to Predict Performance Versus Design

To enable complete closure of a Sabatier-based atmosphere revitalization system where oxygen recovery is predicated by carbon dioxide reduction to methane and water, there must be a means of recovering the hydrogen from the byproduct methane. Honeywell’s carbon vapor deposition (CVD) reactor is a potential technology that provides a thermochemical means by which methane decomposes at high temperatures (> 1,000 °C) into hydrogen, solid carbon, and a small percentage of hydrocarbons. This advanced methane pyrolysis reactor utilizes a carbon fiber substrate to capture and immobilize the solid pyrolytic carbon rather than allowing continuation toward undesirable soot. The CVD reactor is a highly dynamic system whose characteristic geometry changes while operating. The reactor model was developed to provide predictive performance assessment capabilities at different operating conditions, geometries, and configurations to aid in technology development. Modeling of the CVD reactor considered the multitude of simultaneous and interdependent physicochemical processes that are entailed, including: (1) the homogeneous gas-phase pyrolysis reactions, (2) the heterogeneous deposition reactions, and (3) the transport of gaseous species in the reactor domains. A 1-D CVD reactor model was developed based on a general packed-bed, axial-dispersion plug flow reactor model with a solid substrate balance to account for carbon deposition. The model includes 59 species and 243 reactions constituting the methane pyrolysis reaction network, carbon deposition, and interconversions of oxygenate species. The reactor model was compared to brassboard test data, where it was shown to be able to capture the pertinent physicochemical phenomena as evidenced by its agreement with short-duration run data (i.e., looking at different temperature and flow rate conditions), as well as trend agreement with the brassboard reactor’s long-duration test data. Recommendations are provided on further model development steps as well as its use cases.

Air Revitalization↗
DOE OSTI2017

Preliminary Study for Development of Technical Basis of Quantitative Safety Goals for NPPs Operations in UAE

In this study, frameworks and methodologies for the development of technical basis of quantitative safety goals will be established. Most studies on the technical basis of the quantitative safety goals have been conducted after the commercial operations of NPPs. As a matter of fact, PSA insights are of much use for these kind of studies, hence it was not practical to provide the technical bases in a timely manner due to the unavailability of PSA studies at the moment. FANR has adopted the following mean values as quantitative safety goals: overall core damage frequency (CDF) < 10{sup -5}/ry and overall large release frequency (LRF) < 10{sup -6}/ry. Capabilities, framework, and methodologies to develop the technical basis of the quantitative safety goals in the UAE are established. Moreover, preliminary technical bases are developed in an early phase of NPP program in the UAE, which will lead to establishment, management, and update of the quantitative safety goals in a later phase.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Molten Chloride Reactor Experiment

MCRE will be the first fast spectrum, molten chloride reactor in operation. Small critical experiment to provide data for future reactors. - Collaboration between Southern Company Services, TerraPower, and Idaho National Laboratory (INL), as well as other partners. - Expected to be built and operated at the LOTUS testbed (ZPPR). - Authorized by the U.S. Department of Energy. - Technology development of salt synthesis.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Spurious actuations in digital instrumentation and control systems - evaluation framework - 137

When a digital Instrumentation and Control (I and C) system or its associated components produce an unintended operation, it is known as a spurious actuation. Spurious actuations could lead to unnecessary challenges to safety equipment, challenge the ability of safety systems to provide their intended functions, or place the plant in an un-analyzed state. There is arguably a lack of clear and sufficient regulatory guidance for assessing spurious actuations. In an attempt to address this guidance gap, this paper provides a generic framework for evaluating spurious actuations in digital I and C systems, components, or supporting systems that are important to safety. The framework provides a methodology for: (1) defining the scope of the evaluation including supporting assumptions; (2) providing options for excluding a spurious actuation from the evaluation; (3) assessing the potential consequences from the assessed spurious actuations; and, (4) defining a high-level acceptance criteria. The methodology discussed by this paper is not to be construed as a requirement, regulation, or acceptable guidance by either domestic or international regulators; instead, it is intended to serve as a potential foundation or technical basis to be used for developing clear and sufficient regulatory guidance for assessing spurious actuations in digital I and C systems. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

A novel priority selection system for nuclear power plant - 287

The priority selection system is an important component of security level instrument control platform for nuclear power plant, the main function of which is priority processing the manual or automatic command with various security levels and driving engineering safety features and field devices, such as pump and valve. Its stability and safety plays a crucial role in the operation of the power plant. The main objective of the present paper is to describe the new nuclear power plant-oriented priority selection system, which is developed independently by China Nuclear Control System Engineering Company. In this system, the control commands from 1E DCS, ECP and DAS systems are transmitted via hardwire, while those from NC DCS are transmitted by field-bus. The control commands experience priority output to control the field actuators. Meanwhile, in order to control key field devices when failure of the instrument control systems occurs, the issuing interface for native instructions with permissions setting is also offered. The whole system provides collection-feedback functions for field devices, high coverage self-diagnostic functions and periodic test functions for safety systems. Moreover, the feedback information of field devices, the results of self-diagnosis and periodic test will be submitted to DCS of nuclear power plant. In addition, the core part of priority selection system, namely priority selection module, is obtained by pure FPGA technology, which is different from DCS systems and could avoid common cause failure of software effectively. (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 OSTI2025

Build-to-Replace Strategy to Reduce Operations and Maintenance Costs for Advanced Reactors

This paper targets the reduction of fixed operations and maintenance (O&M) costs for advanced reactor (AR) designs. This goal will be severely constrained if the underlying assumptions of O&M approaches and practices are not questioned and reexamined, especially today when changes can be implemented effectively and efficiently. Achieving a reduction in AR O&M costs requires a completely new way of thinking -a paradigm shift- not through incremental, technology-focused approaches alone. Here, we address this challenge by evaluating the impact of moving to a shorter design life for major structures, systems, and components and shorter, more predictable refurbishment cycles as modeled by the commercial airline industry: the build-to-replace approach. This paper provides a brief overview of this different mindset to O&M applied to ARs and it provided a set of analytical tools based on multi-objective optimization to identify the benefits of such an approach. In conclusion, we provide a direct example analysis of a build-to-replace scenario by identifying and evaluating scenarios for reduced system and component lifetimes and associated replacement and refurbishment schedules to evaluate impacts on O&M costs and other lifecycle elements, such as structure, system, and component reliability.

97 - MATHEMATICS AND COMPUTING↗
DOE OSTI2025

Simulation Results of a Thermal Power Dispatch System from a Generic Pressurized Water Reactor in Normal and Abnormal Operating Conditions

Amid economic pressures in the U.S. electricity market, nuclear utilities are exploring new revenue streams, including hydrogen production. A generic pressurized water reactor simulator was modified to incorporate a novel design for a TPD system coupled to a hydrogen production plant. Standard malfunctions were included in the simulation design, including steam line breaks at various system locations and flow interruptions in the hydrogen plant due to multiple faults, reflecting anticipated operational challenges. It is imperative that the TPD system operation has a minimal effect on the reactor power, primary coolant system, and turbine system operation and performance. Due to the specific design and application of this TPD system, with the proposed turbine control system changes, the overall impact on the existing plant systems is low. Normal TPD operating scenarios resulted in minor effects on the existing plant systems: reactor power changes by at most 0.2%, and gross generator output changes by 20.5 MWe from 100 MWt of TPD. The most severe malfunction analyzed in this work is a full TPD steam line break downstream of the extraction location, which results in an increase in reactor power of about 0.5%. The gross generator output decreases by 36 MWe, a total decrease of 60 MWe from the full power steady state (FPSS) condition. These results indicate that an industrial hydrogen production plant could be coupled thermally to a nuclear power plant with limited effects on the existing system operation and safety.

08 HYDROGEN↗
DOE OSTI2024

A Framework to Assess Advanced Reactor Spent Fuel Management Facility Deployment

Advanced nuclear reactors offer various operational advantages over existing light water reactors but could produce types of spent nuclear fuel (SNF) with a wide variety of forms and characteristics depending on how many different concepts are deployed. Each advanced reactor SNF type potentially poses unique management challenges. New planning efforts will be necessary to anticipate how the management requirements of advanced reactor SNF will affect the deployment of an integrated waste management system. This paper applies a framework of high-level facility deployment milestones to a generic SNF management system, reviewing them together with the advanced reactor SNF characteristics and management requirements. This allows for the investigation of factors that influence facility and system deployment, and ultimately, the identification of challenges facing the deployment of different kinds of SNF management facilities. Here, the back end of the once-through fuel cycle is examined for four advanced reactor system technology types: sodium-cooled fast reactors, high-temperature gas-cooled reactors, liquid-fuel molten salt reactors, and lead-cooled fast reactors. It is observed that milestones earlier in the facility deployment process (e.g., siting and facility design) are more impacted by the uniqueness of advanced reactor SNF characteristics than others (e.g., construction and testing). Ultimately, none of the differences are seen as fundamentally disqualifying in a technical sense; however, they should be considered early, potentially as part of reactor design, to avoid issues in the future.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Pressure Composition Temperature Curves of Hydride Moderators

Microreactors are compact and mobile reactors which operate at low power and high temperatures. These reactors generate heat which is then converted to electricity for use in a variety of special purpose applications. Microreactors are of interest due to their light weight, compact size and the ability to meet the needs for high energy in difficult or remote areas. Metal hydrides have been studied for nuclear and aviation applications. The emergence of microreactors the interest in metal hydrides is significant due to their ability to moderate neutrons in a light weight and solid form. In case of an off normal event, more accurate data is needed for the absorption/desorption rates of the hydrides due to the high uncertainties on hydrogen redistribution in the moderator. The literature shows considerable discrepancies in the PCT curves for yttrium and zirconium. The uncertainty is much more notable at higher stoichiometries of the hydride phase at high temperature. To understand the in-pile performance, a detailed understating of the impurities and microstructure is needed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗