Toward a Big Data-Based Approach: A Review on Degradation Models for Prognosis of Critical Infrastructure

2020 ◽  
Vol 4 (2) ◽  
Author(s):  
Guru Prakash ◽  
Xian-Xun Yuan ◽  
Budhaditya Hazra ◽  
Daijiro Mizutani
Keyword(s):  
Big Data ◽  
Nuclear Power ◽  
Power Plants ◽  
Critical Infrastructure ◽  
Degradation Mechanism ◽  
Modern Society ◽  
Small Data ◽  
Practical Applications ◽  
Degradation Modeling ◽  
Degradation Models

Abstract Safety and reliability of large critical infrastructure such as long-span bridges, high-rise buildings, nuclear power plants, high-voltage transmission towers, rotating machinery, and so on, are important for a modern society. Research on reliability and safety analysis started with a “small data” problem dealing with relative scarce lifetime or failure data. Later, degradation modeling that uses performance deterioration, or, condition data collected from in-service inspections or online health monitoring became an important tool for reliability prediction and maintenance planning of highly reliable engineering systems. Over the past decades, a large number of degradation models have been developed to characterize and quantify the underlying degradation mechanism using direct and indirect measurements. Recent advancements in artificial intelligence, remote sensing, big data analytics, and Internet of things are making far-reaching impacts on almost every aspect of our lives. The effect of these changes on the degradation modeling, prognosis, and safety management is interesting questions to explore. This paper presents a comprehensive, forward-looking review of the various degradation models and their practical applications to damage prognosis and management of critical infrastructure. The degradation models are classified into four categories: physics-based, knowledge-based, data-driven, and hybrid approaches.

Strategies for Treating Weld Residual Stresses in Probabilistic Fracture Mechanics Codes

2011 ◽  
Author(s):  
Frederick W. Brust ◽  
R. E. Kurth ◽  
D. J. Shim ◽  
David Rudland
Keyword(s):  
Fracture Mechanics ◽  
Residual Stresses ◽  
Stress Corrosion ◽  
Stress Corrosion Cracking ◽  
Nuclear Power ◽  
Power Plants ◽  
Degradation Mechanism ◽  
Welding Process ◽  
Corrosion Cracking ◽  
Uncertainty Sources

Risk based treatment of degradation and fracture in nuclear power plants has emerged as an important topic in recent years. One degradation mechanism of concern is stress corrosion cracking. Stress corrosion cracking is strongly driven by the weld residual stresses (WRS) which develop in nozzles and piping from the welding process. The weld residual stresses can have a large uncertainty associated with them. This uncertainty is caused by many sources including material property variations of base and welds metal, weld sequencing, weld repairs, weld process method, and heat inputs. Moreover, often mitigation procedures are used to correct a problem in an existing plant, which also leads to uncertainty in the WRS fields. The WRS fields are often input to probabilistic codes from weld modeling analyses. Thus another source of uncertainty is represented by the accuracy of the predictions compared with a limited set of measurements. Within the framework of a probabilistic degradation and fracture mechanics code these uncertainties must all be accounted for properly. Here we summarize several possibilities for properly accounting for the uncertainty inherent in the WRS fields. Several examples are shown which illustrate ranges where these treatments work well and ranges where improvement is needed. In addition, we propose a new method for consideration. This method consists of including the uncertainty sources within the WRS fields and tabulating them within tables which are then sampled during the probabilistic realization. Several variations of this process are also discussed. Several examples illustrating the procedures are presented.

scholarly journals Human factor in emergency occurrence at NPP during the pandemic COVID-19: new potential risks and recommendations to minimize them

2021 ◽  
Vol 280 ◽  
pp. 09013
Author(s):  
Oleksandr Popov ◽  
Anna Iatsyshyn ◽  
Norbert Molitor ◽  
Andrii Iatsyshyn ◽  
Yevhen Romanenko ◽  
...  
Keyword(s):  
Mental Health ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Critical Infrastructure ◽  
Well Being ◽  
Psychological Support ◽  
Negative Consequences ◽  
Large Numbers ◽  
Potential Risks

COVID-19 pandemic is an emergency that spread across the planet. It threatens lives and health of the population, with many dead and injured. Also, during pandemic there was (and continues to be) a violation of normal living conditions, significant material damage to economies of different countries. Quarantine restrictions affected social, educational, cultural and spiritual spheres of society. There is an urgent threat to operation of critical infrastructure as the pandemic led to large numbers of hospitalized or self-isolated people and being treated at home. At present, the operation of critical infrastructure, and in particular nuclear power plants, is strategically important for the proper functioning of any country. Also, the research describes causes of emergencies at nuclear power plants and identifies another factor - the mass disease of NPP personnel or their relatives on COVID-19. It is determined that timely detection and appropriate work to restore mental health during the pandemic is no less important area of work along with providing medical care to population. Stress associated with uncertainty, threat to health, can have negative consequences for well-being and mental health of workers. Depression, emotional exhaustion, anxiety, irritability, suicidal thoughts, etc. can also be consequences. Socio-psychological support especially for critical infrastructure workers is important to promote mental health during COVID-19 pandemic. Recommendations for socio-psychological support of NPP personnel were developed. Specialized mobile applications are described as convenient and economical way for primary psychological care.

Grid Collapse Security, Stability and Vulnerability Issues: Impactful Issues Affecting Nuclear Power Plants, Chemical Plants and Natural Gas Supply Systems

2018 ◽  
Vol 16 (1) ◽  
Author(s):  
Robert McCreight
Keyword(s):  
Natural Gas ◽  
Nuclear Power ◽  
Power Plants ◽  
Critical Infrastructure ◽  
Geomagnetic Storms ◽  
Risk Mitigation ◽  
The United States ◽  
Primary Energy ◽  
Chemical Plants ◽  
Cyber Attack

Abstract Vulnerability to extended power outages stemming from grid collapse triggered by terrorism, technological accident, cyber attack or geomagnetic storms is understood to mean the widest possible spectrum of immediate and downstream consequences for our nations critical infrastructure. Regrettably few realistic plans are in place for dealing with this risk especially as it pertains to three primary energy systems of strategic significance to the United States – nuclear power, chemical manufacturing and natural gas supplies. The author argues that greater sustained attention is needed to upgrade the resilience of these systems, foster greater sharing of remedies among them to offset the worst effects of grid collapse which exceeds 15 consecutive calendar days and build collective avenues of enhanced risk mitigation against such scenarios.

Development of RI-ISI Program for Krško NPP

2013 ◽  
Author(s):  
Scott Kulat ◽  
Robin Graybeal ◽  
Benjamin Montgomery ◽  
Ivan Vrbanić ◽  
Ivica Bašić ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Selection Process ◽  
Degradation Mechanism ◽  
Electrical Power ◽  
Implementation Program ◽  
Class 1 ◽  
Process Risk ◽  
Almost All ◽  
The U.S

Risk-informed methodologies for inservice inspections of safety related piping in nuclear power plants were formally established in mid-1990s in the U.S. Since then, they have been adopted and applied by almost all of the U.S. plants. Nowadays, risk-informed inservice inspection (RI-ISI) is considered to be a standard for the operating plants in the U.S. It was not long before the RI-ISI practice started to be “exported” from the U.S. to other countries. By now, RI-ISI had found its way to a number of European and other countries. Among the recent examples is the Krško Nuclear Power Plant (Krško NPP), a two-loop Westinghouse-designed PWR located in Slovenia and owned by Slovenian and Croatian utilities. Krško NPP finished its third inservice inspection (ISI) interval in July 2012 and initiated implementation of the RI-ISI program at the start of the fourth interval. The process used to develop the RI-ISI program conformed to the methodology described in Electrical Power Research Institute (EPRI) Topical Report TR-112657 and included a degradation mechanism evaluation, consequence analysis and risk characterization for ASME Class 1 and Class 2 piping, as well as an element/examination selection process, risk impact assessment and inspection implementation program development. This paper describes the development of the Krško NPP RI-ISI program and the results of its RI-ISI application. A discussion is, also, provided on some aspects relevant for application of RI-ISI approaches developed in the U.S to plants outside of the U.S.

CONCEPTUAL BIG DATA CASE STUDY TO IDENTIFYING RISKS OF NEW NUCLEAR TECHNOLOGIES

2019 ◽  
Vol 6 (1) ◽  
Author(s):  
Jaeheum Yeon ◽  
Mark Czarny ◽  
John Walewski ◽  
Julian Kang
Keyword(s):  
Big Data ◽  
Content Analysis ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
New Technologies ◽  
Risk Identification ◽  
Visual Content ◽  
Short Period ◽  
Identification Approach

New technologies associated with nuclear power plants are being introduced regularly. However, many of the risks and uncertainties associated with these new nuclear technologies have yet to be identified. In this study, the risks related to newly-developed nuclear technologies were determined through an extensive review of the extant literature. A qualitative visual content analysis was selected as the research method employed to identify words repeatedly occurring in 147 journal articles. Through this conceptual “big data” approach, frequently mentioned words were identified using a co-occurrence map. The analysis results were then grouped into four categories: fuel resources, operational system designs, nuclear reactor cooling systems, and steam generators. Words used repeatedly to reference these four key categories were determined to also represent potential causes of risk factors. Many texts could be analyzed in a short period of time through the use of visual content analysis software. Frequently emphasized correlating words were then identified. This big data approach can also be applied to additional nuclear power practices to identify other uncertainties. Last, the limitations of a visual content analysis employed as a risk identification approach were revealed through this study.

scholarly journals Novel Hydroxyapatite Beads for the Adsorption of Radionuclides from Decommissioned Nuclear Power Plant Sites

2021 ◽  
Vol 11 (4) ◽  
pp. 1746
Author(s):  
Thi Nhung Tran ◽  
Junho Kim ◽  
Joo-Sung Park ◽  
Youngkun Chung ◽  
Jaemun Han ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Polyvinyl Butyral ◽  
Order Kinetic ◽  
Practical Applications ◽  
Adsorption Mechanisms ◽  
Rate Limiting Step ◽  
Aqueous Environments ◽  
Order Kinetic Model ◽  
Pseudo Second Order

Although a powdered form of hydroxyapatite (p-HdA) has been studied for the adsorption of heavy metals that contaminate the restoration sites of decommissioned nuclear power plants, most of the studies are limited in the laboratory due to the head loss and post-separation in practical applications. Herein, we fabricated a porous bead form of HdA (b-HdA) as a novel adsorbent for removing radionuclides from aqueous environments via a facile synthesis by mixing the p-HdA precursor and polyvinyl butyral (PVB) as a binder and added a sintering process for the final production of a porous structure. The spherical b-HdA with an approximate diameter of 2.0 mm was successfully fabricated. The effectiveness of the b-HdA at removing Co(II) was investigated via the adsorption equilibrium at various experimental temperatures. The b-HdA exhibited the adsorption capacity for Co(II) ions with a maximum of 7.73 and 11.35 mg/g at 293 K and 313 K, respectively. The experimental kinetic data were well described using a pseudo-second-order kinetic model, and the adsorption mechanisms of Co(II) onto the b-HdA were revealed to be a chemisorption process with intraparticle diffusion being the rate-limiting step. In addition, the competitive adsorption onto the b-HdA with the order of U(VI) > Co(II) > Ni(II) > Sr(II) > Cs(I) was also observed in the multi-radionuclides system. Considering the advantages of the size, applicability to the continuous-flow column, and the easy separation from treated water, the b-HdA can be an excellent absorbent with high potential for practical applications for removing radionuclides.

Contribution to Assessments of Impacts on the Critical Infrastructure

2014 ◽  
Vol 656 ◽  
pp. 578-587
Author(s):  
David Valis
Keyword(s):  
Risk Management ◽  
Fuzzy Logic ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Critical Infrastructure ◽  
Potential Threat

Assessing the vulnerability of critical infrastructure objects is of major concern when dealing with the process of dependability and risk management. Special attention is paid to the objects of higher interest, such as nuclear power plants. In spite of the protection of these objects, there is still a certain level of a potential threat. The aim of the paper is to describe a possible way of assessments of the consequences when attacking on an object of critical infrastructure. Several characteristics of the impacts outcomes were obtained. For this reason, as well as for modelling of adversary ́s behaviour impact, fuzzy logic tools have been used.

Comparison of ASME Code NB-3200 and NB-3600-Based Environmental Fatigue Calculations for Piping Components

2014 ◽  
Author(s):  
Kyeongjin Yang ◽  
Daesu Kim ◽  
Dongjae Lee ◽  
Joonho Lee ◽  
Sangbae Lee ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Time Consistency ◽  
Correction Factors ◽  
Pressurized Water ◽  
Water Reactor ◽  
Practical Applications ◽  
Thermal Transients ◽  
Asme Code ◽  
Fatigue Evaluation

Environmentally-assisted fatigue evaluations are to be conducted for ASME Code Class 1 piping components in a pressurized water reactor. Environmental fatigue correction factor method for incorporating the effects of light water reactor coolant environments into ASME Section III fatigue evaluations was investigated in this paper. Both ASME Code NB-3200 and NB-3600 methods were used to determine the usage factors of the piping components. Considered in these calculations were the loads which are generally applied to the piping design for the nuclear power plants such as seismic, thermal expansions, thermal transients, thermal stratifications and building-filtered dynamic loadings. For the practical applications of NB-3600 method, regarded as the simple and conservative approach, to the piping components, it was presumed that the stress intensity and/or strain time histories for all or some of the external loadings were not known; therefore the time consistency might not be considered in calculating the usage factors as well as environmental correction factors (Fen). In NB-3200 method in contrast to NB-3600, the stress variations with time for all loads except for the dynamic loads were obtained for the fatigue evaluations in LWR environments, and therefore the time consistency was considered. The results showed that the environmental correction factors as well as in-air cumulative usage factors calculated from NB-3200 methods were significantly less than those from NB-3600 rules. In addition, comparing the results of conventional ASME fatigue evaluation applied until 2006 to the ones in accordance with USNRC RG 1.207 issued on 2007, one may identify that the cumulative usage factors in LWR environments were larger than the conventional one due to the change of design fatigue curves as well as Fen factors accounting for the environmental effects on fatigue. Although this work was focused on the detailed calculations of the usage factors and Fen values, one might identify or suggest a number of areas requiring further clarification or research through the efforts of this study, which were not yet addressed. A few items needed to be clarified, especially for NB-3600-based fatigue evaluations, are also discussed in this paper.

Confluent Modeling of Heterogeneous Safety and Operational Instrumentation and Control Systems

2020 ◽  
Vol 6 (3) ◽  
Author(s):  
Mithil Parekh ◽  
Yuan Gao ◽  
Mariana Jockenhoevel-Barttfeld ◽  
Karl Waedt
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Critical Infrastructure ◽  
Heterogeneous Systems ◽  
Legacy Systems ◽  
Security Measures ◽  
Product Families ◽  
Or Groups ◽  
Specialized Equipment ◽  
Commercial Off The Shelf

Abstract Individual systems or groups of systems related to the safety and operational instrumentation & control (I&C), and electrical systems are gradually replaced in existing nuclear power plants (NPPs) as part of modernization projects. Modernizations are usually scheduled over multiple years. Different automation platform generations and safety-related product families are deployed to progressively replace legacy systems. Typically, each of the new I&C platforms and products have their own set of engineering tools. One challenge for the modernization of installed systems is the safety and security approval of these heterogeneous systems by regulatory bodies. An NPP consists of systems manufactured by different vendors, which target different application domains, like category A according to IEC 61226 for a reactor protection system, category C for some control room data processing systems or nonclassified for auxiliary systems. Thus, replacing a legacy system by introducing a new one might bring in unknown risks, especially when the new system has more complex interfaces, e.g., when replacing an analog system by a digital system. Moreover, considering the time and budget limitations, commercial-off-the-shelf (COTS) hardware and software are also involved in modernization projects. In contrast to specialized equipment, vulnerabilities for COTS are widespread. On the other hand, existing security measures/mitigations are also required to reflect the system's changes, e.g., mitigations for known vulnerabilities of COTS systems. In order to make an overall and integrated safety analysis after a system change in the frame of a modernization project, it is necessary to jointly consider these I&C systems targeting different application domains, in addition to the physical aggregates, like sensors, pumps, and valves that interact with the physical processes. The restrictive deployment of wireless technologies may also be modeled and analyzed. While wireless is not deployed by legacy systems, it is being covered by new nuclear IEC standards, as some utilities intend to simplify selected I&C maintenance procedures that involve temporary data collection. The key modeling concepts consider new developments in the critical infrastructure and industrial automation domain. With the integrated modeling approach, different disciplines can be addressed, like probabilistic and deterministic safety analyses, security assessments, need for testing and specialized trainings.

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