Systems, And Components Research Articles

A risk-informed and performance-based approach to defining prevention and mitigation

The terms “prevention” and “mitigation” are often used to define safety functions in nuclear power plants and other complex engineered systems. For existing light water reactor (LWR) nuclear plants, these terms are typically used in the context of preventing core damage and mitigating the consequences of a severe accident. Balancing the reliance on prevention and mitigation is often stated as an important principle of defense-in-depth. To apply these terms to formulate design requirements for advanced reactors, it is necessary to define specifically what is to be prevented and mitigated because the concept of core damage is not generally applicable to advanced non-LWR concepts and designs. Advanced non-LWR reactor concepts use different fuels, moderators, coolants and different strategies for achieving the functional containment of radionuclides. The adoption of the small modular reactor concept also leads to plants with many small reactors. This paper discusses a risk-informed and performance-based definition of prevention and mitigation that can be applied to any reactor concept or design. It was developed as part of the Licensing Modernization Project (LMP) which has introduced a new approach to developing a safety case to support design and licensing. A new way of thinking about prevention and mitigation was needed to formulate advanced reactor design and special treatment requirements for structures, systems and components (SSCs) that could be applied to a wide diversity of designs. The LMP approach discussed in this paper includes criteria to identify safety significant SSC prevention and mitigation functions and uses SSC reliability and capability targets to inform the selection of design and special treatment requirements and to support the evaluation of defense-in-depth adequacy. These prevention and mitigation functions are defined in the context of event sequences that comprise the licensing basis events that anchor the safety case for the reactor operating license.

Nuclear Facilities Maintenance Culture Development Study

Maintenance is an essential part of the assets and facilities in the organization/company. This deserves serious attention because it is necessary for organizational/company resources, especially for aging nuclear facilities/installations. Maintenance is the main program used to manage company assets and facilities. A lack of asset and facility maintenance culture increases management and maintenance costs. A maintenance culture is needed to improve maintenance skills, tenacity, and perseverance. Maintenance culture is recognized as an essential aspect to improve the quality of maintenance work to maintain structures, systems, and components (SSC) so that they are ready for use at any time/when needed, as well as part of the implementation of aging management programs to extend the life of assets and facilities. This paper will discuss the development of the maintenance culture and review poor maintenance cultures and their solutions. This paper aims to develop a maintenance culture to prevent premature aging and extend the life of assets and facilities. Factors developing maintenance culture include leadership, communication, rewards and recognition, responsibility, rules and standards, empowerment, motivation, involvement, policies, work strategy and planning, teamwork, training and education, awareness, and organizational structure. Corruption, poor leadership, attitude problems, and lack of policy cause poor maintenance culture. The solution to a poor maintenance culture is selecting the right people, improving job skills and knowledge, repairing equipment, ensuring machine/equipment readiness, good planning, scheduling, and general organization, measuring performance, motivating and encouraging desire, and providing sufficient funding. Maintenance culture has an impact on facility performance. Keywords: maintenance culture, nuclear organization, assets, nuclear facility

Probabilistic typhoon hazard and sensitivity analysis for nuclear power plant sites in Korea using logic tree

There has been an increase in typhoon intensity, and high-intensity typhoons can affect the safety of nuclear power plants (NPPs). Typhoons can damage the structures, systems, and components (SSCs) of NPPs, causing core damage and the release of radioactive materials into the environment. Therefore, design criteria for the effects of natural phenomena are demanded to ensure the safety-critical SSC functions of NPPs. Also, there is a need for safety assessments of typhoons in operational NPPs, which require typhoon hazard analysis. In this study, we derived typhoon hazards for NPPs in Korea using Logic Tree and Monte Carlo simulations. We performed sensitivity analysis on the typhoon hazards based on the distances (up to 100 km) from the NPP sites. Among other NPP sites, the typhoon hazard of the Kori site showed the highest wind speed, and the Hanul site showed the lowest wind speed. The uncertainty of the wind speed increased as the recurrence interval of the typhoon hazard increased. We compared the typhoon hazard in the areas of interest to that within 100 km. The range of the distance deviation for the Hanbit site's typhoon hazard was within μ±σ up to 75 km, while for the other sites, it was within μ±σ up to 100 km.

Reviewing Welding Procedures—Checklists for Nuclear Power Systems

Abstract To ensure effective control of the welding process, nuclear power architecture and engineering (AE) companies routinely undertake the review of procedure qualification records (PQRs) and welding procedure specifications (WPSs) for safety-related structures, systems, and components (SSCs), as qualified by manufacturers and construction contractors. Given the substantial quantity of PQRs and WPSs necessitating evaluation, along with the multitude of variables inherent in these documents, a set of review checklists has been devised. These checklists, developed from the perspective of AE companies' welding engineers, transform welding procedure qualification rules and requirements into intuitive descriptions, assessments of correctness, or numerical comparisons of scale. AE companies' welding engineers employ these review checklists to scrutinize PQRs and WPSs, facilitating a comprehensive, accurate, and efficient review process.

A comprehensive framework for identifying safety-related and ageing sensitive sscs in low power nuclear research reactor

This work aims to establish a novel approach for the selection of structures, systems, and components (SSCs) important to safety and vulnerable to ageing degradation for low-power research reactors. For this purpose, the framework developed combines the Preliminary Risk Analysis (PRA) technique to categorize SSCs into four different safety classes, and Ageing Failure Mode and Effect Analysis (AFMEA) technique to find out ageing failure modes related to a specific SSC and evaluate them according to the Risk Priory Number (RPN). The intersection of the PRA and AFMEA outcomes then yields a list of SSCs that are both important to safety and strongly ageing sensitive, making it possible to optimize the selection process of SSCs to be prioritized in further preventive maintenance, periodic testing, and inspection activities. The present paper also conducts a case study to apply the proposed methodology to the Brazilian IPR-R1 TRIGA nuclear research reactor. The selected SSCs were then both AL and SS fuel rods from the reactor core, stainless steel pipe, thermocouples, pressure indicators, and water level indicators from the primary circuit, and power monitoring channels, radiation monitors, and cooling water temperature channels from the controlling system. Moreover, the approach was also able to identify SSCs that are important from a safety perspective but are not ageing sensitive (reactor tank and control rods) and components that are not important to safety but are strongly ageing sensitive (pipe filters). These findings may further serve as input for the comprehensive safety assessment conducted during the Periodic Safety Review of the facility.

Ageing of Nuclear Power Plant’s Equipment. Assessment of Reactor Pressure Vessel Ageing Effect

Ageing of materials stands for the change in their mechanical properties, due to thermodynamic imbalance of the initial condition, and gradually bringing the structure to equilibrium in the presence of sufficient diffusive mobility of the atoms. The current paper presents the main mechanisms of degradation of the mechanical properties of the nuclear power plant equipment metal, as well the models for ageing. The ageing effects and ageing indicators typical for the equipment are defined. The ageing process may cause loss of the appropriate functions. It is important to recognize the ageing effects, as well the ageing indicators. Testing methods are defined for every ageing indicator. The paper considers different national approaches to ageing management of structures, systems, and components (SSCs) at nuclear power plants (NPPs).

Proposed Enhancements to the Risk-Informed and Performance-Based Regulatory Framework for Seismic Hazard Design at NRC-Regulated Nuclear Power Plants

The commercial nuclear power plant industry initiated the licensing modernization project (LMP) to enhance the risk-informed and performance-based (RIPB) regulatory basis for advanced nuclear power reactors. The LMP framework relies heavily on RIPB concepts and approaches that together integrate the defense-in-depth philosophy. One example approach for seismic design is to align the LMP concepts with the performance targets described in the American Society of Civil Engineers (ASCE) standard, ASCE 43-19. The underlying strategy of this approach is to consider the performance of individual structures, systems, and components (SSCs) in seismic design, as well as the role they play in an accident event sequence. This approach contrasts with current regulations, in which every individual safety-related SSC is designed to the same seismic criteria irrespective of the role the SSC plays in the overall system performance. This new philosophy envisions more flexible seismic design options for each SSC, such that the overall seismic design can meet system-level acceptability criteria as well as plant-level acceptability criteria. The objective of this paper is to illustrate the flexibility and benefits of this proposed approach to the seismic design of SSCs in terms of reduced SSC demands (by reducing the design ground motions for SSCs) and improved SSC capacities (by allowing for alternative damage state limits). A simple shear wall was designed using ASCE 43-19 for a hard rock site in the Central Eastern United States considering alternate seismic design category and limit state combinations to examine the physical designs and functional fragilities of these combinations and their impact on seismic performance. The flexibility of this proposed approach is illustrated by an example that shows reduced SSC demands, while the SSC capacities and margins remain consistent with the required safety performance without any loss in overall plant safety.

A Future with Machine Learning: Review of Condition Assessment of Structures and Mechanical Systems in Nuclear Facilities

The nuclear industry is exploring applications of Artificial Intelligence (AI), including autonomous control and management of reactors and components. A condition assessment framework that utilizes AI and sensor data is an important part of such an autonomous control system. A nuclear power plant has various structures, systems, and components (SSCs) such as piping-equipment that carries coolant to the reactor. Piping systems can degrade over time because of flow-accelerated corrosion and erosion. Any cracks and leakages can cause loss of coolant accident (LOCA). The current industry standards for conducting maintenance of vital SSCs can be time and cost-intensive. AI can play a greater role in the condition assessment and can be extended to recognize concrete degradation (chloride-induced damage and alkali–silica reaction) before cracks develop. This paper reviews developments in condition assessment and AI applications of structural and mechanical systems. The applicability of existing techniques to nuclear systems is somewhat limited because its response requires characterization of high and low-frequency vibration modes, whereas previous studies focus on systems where a single vibration mode can define the degraded state. Data assimilation and storage is another challenging aspect of autonomous control. Advances in AI and data mining world can help to address these challenges.

Research to Develop Flood Barrier Testing Strategies for Nuclear Power Plants

The U.S. Nuclear Regulatory Commission has developed regulations regarding the siting and design of nuclear power plants (NPPs) that are aimed at addressing various natural hazards, including flooding. Flood barriers are designed to prevent water from entering NPP areas containing structures, systems, and components (SSCs) important to safety. The barriers are used at NPPs along with drains, sumps, pumps, valves, plugs, and site grading as part of the plant flood protection features that protect SSCs from experiencing external or internal flooding and mitigate the effects of flooding on NPP operations. The performance of flood protection features, including flood barriers at NPPs, has been an ongoing concern. Domestic and international operational experience provides clear indications that flood barrier performance has significant safety implications, especially for aging NPPs. The observed deficiencies show that flood barriers should be designed and installed properly, then adequately tested, inspected, and maintained in order to ensure that they perform their intended functions during flooding events. This paper reviews available information related to flood barriers employed at U.S. NPPs and provides an overview and categorization of NPP flood barriers. It identifies potential domestic and international flood barrier testing facilities, including operating and decommissioned U.S. NPPs. Finally, this paper presents the technical and logistical considerations that should be made when developing specific testing strategies and protocols for flood barriers, such as the selection of flood barriers, test locations, testing approach, performance criteria, and testing parameters.

Bayesian Network–Based Fault Diagnostic System for Nuclear Power Plant Assets

Future advances in nuclear power technologies call for enhanced operator advice and autonomous control capabilities that can leverage simpler designs and increased safety features to reduce reliance on human labor. One of the first tasks in the development of such capabilities is the formulation of symptom-based conditional failure probabilities for the plant structures, systems, and components (SSCs) of interest. The primary goal is to aid plant personnel in (1) deducing the probabilistic performance status of the monitored SSCs and (2) detecting impending faults/failures. The task of estimating conditional failure probability is a bidirectional inference problem, and a logical approach is to use the Bayesian network (BN) method. As a knowledge-based explainable artificial intelligence tool and a probabilistic graphical model, BN offers reasoning capability under uncertainty, graphical representation emulating physical behavior of the target SSC, and interpretability of the model structure and results. This paper provides a systematic overview of the BN technique and the software tools for implementing BN models, along with the associated knowledge representation and reasoning paradigm. Both operational data and expert judgment can be readily incorporated into the knowledge base of a BN model. The challenges with data availability are highlighted, and the general approach to target SSC identification is presented. The focus is on failure-prone and risk-important balance of plant assets, especially for cases with strong operator involvement. Two example case studies on the failure of (1) a centrifugal pump and (2) an electric motor are conducted to demonstrate the usefulness and technical feasibility of the proposed BN-based fault diagnostic system using an expert system shell.

Sensitivity analysis of failure correlation between structures, systems, and components on system risk

A seismic event caused an accident at the Fukushima Nuclear Power Plant, which further resulted in simultaneous accidents at several units. Consequently, this incident has aroused great interest in the safety of nuclear power plants worldwide. A reasonable safety evaluation of such an external event should appropriately consider the correlation between SSCs (structures, systems, and components) and the probability of failure. However, a probabilistic safety assessment in current nuclear industries is performed conservatively, assuming that the failure correlation between SSCs is independent or completely dependent. This is an extreme assumption; a reasonable risk can be calculated, or risk-based decision-making can be conducted only when the appropriate failure correlation between SSCs is considered. Thus, this study analyzed the effect of the failure correlation of SSCs on the safety of the system to realize rational safety assessment and decision-making. Consequently, the impact on the system differs according to the size of the failure probability of the SSCs and the AND and OR conditions.

Evaluation of “Safety Related” and “Important to Safety” terminology for safety classification of nuclear installation items in Brazil

In general terms, safety demonstration of nuclear installations is carried out through an assessment of compliance with design criteria and safety requirements established in national and international codes and standards applicable to each type of installation. In addition, a safety analysis consisting of installation behavior study during its useful lifetime, shall be developed considering normal operating conditions, transients, and postulated accidents, to determine safety margins and verify the adequacy of items designed to prevent accidents or mitigate their consequences. Also, design requirements applicable to each installation item depend on its classification with respect to safety. Thus, safety classification of structures, systems, and components (SSCs) must be performed based on adequate methods and clear and consistent criteria to ensure that an overall safety level expected for the installation is achieved. It is worth emphasizing the importance of the terminology adopted and the understanding of concepts definitions used in a safety classification process. The objective of this paper is to present a review of the application of “safety related item” and “item important to safety” terminology, evaluating definitions and interpretations given by the International Atomic Energy Agency (IAEA), the United States Nuclear Regulatory Commission (U.S.NRC) and the National Nuclear Energy Commission (CNEN) of Brazil. In this work, this subject is raised to demonstrate that divergent definitions and misinterpretations of concepts may result in inconsistencies in SSCs safety classification.

Development of classification criteria for non-reactor nuclear facilities in Korea

Non-reactor nuclear facilities are increasing remarkably in Korea combined with advanced technologies such as life and space engineering, and the diversification of the nuclear industry. However, the absence of a basic classification guideline related to the design of non-reactor nuclear facilities has created confusion whenever related projects are carried out. In this paper, related domestic and international technical guidelines are reviewed to present the classification criteria of non-reactor nuclear facilities in Korea. Based on these criteria, the classification of structures, systems and components (SSCs) for safety controls is presented. Using the presented classification criteria, classification of a hot cell facility, a representative non-reactor nuclear facility, was performed. As a result of the classification, the hot cell facility is classified as the hazard category 3, accordingly, the safety class was classified as non-nuclear safety, the seismic category as non-seismic (RW-IIb), and the quality class as manufacturers’ standards (S).

Technical Evaluation Method of Mechanical Equipment Item Substitution in the Procurement and Construction Process of Nuclear Power Plant

Abstract The needs of item substitution are frequent during the procurement and construction of nuclear power project. But the documents prepared for item substitution requests vary considerably in quality that influence working efficiency and project progress. Analyses have been performed for various item substitutions in accordance with laws, regulations, standards, and engineering practices. Focusing on mechanical equipment item substitution, the relative evaluation methods during procurement and construction have been developed based on lessons learned from practices during the engineering practices of advanced passive pressurized water reactor (APPWR) built in China. The methodology of item substitution evaluation can be summarized into six steps, which are critical characteristics for design, nonspecific/specific item substitution, analysis of item function(s) and failure mode(s), impact on related structures, systems and components (SSCs), compensation and coordination measure, and engineering feasibility analysis, respectively. For nuclear power units having been completed, typical cases of item substitution are suggested to be accumulated, and lists of typical critical characteristics for design for the items that often need to be substituted are suggested to be summarized. And then an application library including cases and lists will be established, which will guide item substitution work high quality and efficiency.

Necessity of Coordinating Nondestructive Examination Requirement for Nuclear Power Plant

Abstract This article covers the design management of nondestructive examination (NDE) requirements for nuclear power plants. In recent years, there are many problems related to NDE requirements during the construction of nuclear power plants in China. These problems usually occurred at different stages of the projects, or among the boundaries of different engineering disciplines, which are considered to be related to the incoordination of design management. Under some circumstances, NDE requirements may be repeated, missing or conflicted in different design documents, which leads to inconvenience, confusion and even misdirection to technicians, and increases the risk of omitted inspection and even leads to the occurrence of omitted inspection. Some of these problems have an adverse impact on the quality or progress of projects, and probably also pose potential risks to nuclear safety. In order to avoid these NDE problems that should not have occurred, the owners and designers of nuclear power plant need to pay close attention to promoting the coordination of NDE requirements. Suggestions are made from three aspects: (1) Coordinate the NDE requirements among different class structures, systems and components (SSCs). (2) Coordinate the NDE requirements at the phases of manufacturing, installation and preservice inspection. (3) Standardize the presentation mode of NDE requirements in design documents.

Integrated design of tokamak building concepts including ex-vessel maintenance

The EU-DEMO (referred to as DEMO) tokamak complex currently consists of three buildings, like in ITER: the tokamak building, the tritium building and the diagnostics building. The tokamak building houses the tokamak itself and the numerous plant systems that interface with the necessary systems to produce and control the plasma. It is designed to permit assembly, operation and maintenance of the DEMO tokamak. The vacuum vessel is organized in 8 sectors, with radial ports at the lower and equatorial levels and one vertical upper port. The general architectural structure is arranged around the tokamak with a cylindrical bioshield of 2m thickness around the cryostat and floor levels corresponding to the cryostat penetrations. Additional levels are used for the integration of auxiliary equipment for the various plant systems and for accident mitigation systems. Currently, the tokamak building also represents the final nuclear confinement barrier for the radioactive material towards the environment and the public. This safety function requires the plant and safety systems to limit the release of radioactive substances during normal operation and in accidental conditions well below the safety limits. The complexity of a fusion power plant, like DEMO, with regards to integration of plant systems is much higher than that in a fission power plant due to the larger number of plant systems. Three main criteria drive the integration work of the plant systems inside the tokamak building: (i) safety requirements, (ii) functional requirements of the plant systems themselves and (iii) the maintenance approach. Cost considerations are also taken into account together with the normal and accidental environmental conditions in the various areas of the tokamak building that might challenge the qualifications of structures, systems and components (SSC). The layout of the tokamak building has to be further developed in the Concept Design Phase to follow the plant design evolution providing feedback to the various designers in order to assure that DEMO meets the cited design criteria with an optimized and licensable layout of the most complex nuclear building.

Overview of Nuclear Piping Seismic Verification

Abstract Nuclear power has gone through several transitional phases worldwide with early development, design, and construction in 1960s,1970s, and 1980s, particularly in Japan and US. There were some major research programs conducted during those years to develop methodology and prove the technology. Much of this knowledge is still very relevant and applicable and would be useful to young engineers and new professional coming to the nuclear industry. One of such programs is seismic verification of major structures, systems, and components (SSCs) by tests in Japan during the period of 1981 through 2004 at a large shaking table facility. Opportunities for conducting such large tests are not likely in the current environment of the industry and, therefore, one of the objectives of this paper is to summarize some earlier tests to raise awareness for the newcomers to the industry. This paper discusses on Seismic verification tests of PWR and BWR Primary Loop, Main Steam and Feed Water line, Ultimate Strength of Piping and eroded piping , conducted in Japan. In particular, this paper focuses on: insights obtained from the test results; comparisons of test results with responses during actual earthquakes; collaboration of Japan and US in these tests; observations on robustness of Japanese seismic design; how the results relate to current issues; and use of these results from the future perspectives.

Deterministic and Probabilistic Evaluations of Structures and Components Credited for Seismic Design Extension Conditions

Abstract There is “high confidence” in the ability of the structures, systems, and components (SSCs) of nuclear power plants (NPPs) to perform as designed for design basis accidents (DBAs). For design extension conditions (DECs), the SSCs are required to perform as designed with “reasonably high confidence.” DECs represent scenarios or accidents that are more severe than DBAs. Typically, a spectrum of initiating events including random failure of systems or components, internal and external hazards is used in defining scenarios leading to the DECs. Seismic events that exceed the design basis earthquake (DBE) of a station could be considered seismic DECs. A deterministic design method is proposed to address higher demands of seismic DECs in the new and existing Canada Deuterium Uranium (CANDU) NPPs. The deterministic method builds on the current requirements of applicable codes and standards and recommends more relaxed acceptance criteria. Nevertheless, a means to probabilistically evaluate built-in margin exceeding the demand induced by a seismic DEC would provide a measure of the confidence in a DEC-assigned structure or component performing its function. Therefore, a probabilistic method that estimates the probability of survivability for a structure or component when subjected to the demand induced by a seismic DEC is proposed. The probabilistic method could be used to indicate whether there is a need for applying design modification to existing design features to address demands of seismic DEC. The mean, 5th-percentile, and 95th-percentile fragility functions of these SSCs are used. These fragility functions are typically developed to determine the high-confidence-low-probability-of-failure (HCLPF) value associated with the contribution of a structure or component to the overall plant seismic risk. Sample cases for design features that were implemented in existing CANDU NPPs to address seismic DECs are presented. Both the deterministic and probabilistic methods are applied to cases of civil structures, passive mechanical and electrical components, as well as active control and instrumentation components.

Jet impingement in high-energy piping systems, part I: Characteristics and model evaluation

Under postulated pipe rupture accidents in nuclear power plants (Loss-of-Coolant Accident or LOCA), jet impingement can occur where the high-energy fluid is discharged and cause damage to the surrounding structures, systems, and components (SSCs). As a part of addressing the potential non-conservatisms in the Standard model ANSI/ANS-58.2 (1988), the current work investigates characteristics of jet impingement in high energy piping systems as in nuclear power plants including pressure distribution within jets, pressure on target, and critical mass flux and performs model evaluation. The pressure prediction using ANSI/ANS-58.2 (1988) as well as critical mass flux models are evaluated using the experimental data available in literature. An organized experimental database with different inlet fluid conditions including subcooled water, saturated water/two-phase (0 ≤ x < 0.7), and saturated steam (x ≥ 0.7) is compiled for data analysis and model evaluation. It is found that a power law can be used to correlate the dimensionless static and stagnation pressures at the center of the free jets, while either a power law or an exponential function can be used to correlate the dimensionless stagnation pressure at the center of the targets in impinging jets. The difference in stagnation pressure between free and impinging jets is negligible for saturated steam and saturated water/two-phase jets, but it is significant for subcooled water jets. From model evaluations, the stagnation pressure within the saturated steam jets can generally be predicted well by the Standard model. Although stagnation pressures at far distances (z* > 3.3) can be underestimated, the dimensionless pressures at these axial locations are below 0.07. For saturated water/two-phase jets, the stagnation pressure can be predicted well using the Standard model for the large-scale test data (D > 0.28 m) but are underestimated at some axial locations for the medium- and small-scale test data (D < 0.15 m). Significant overestimation in stagnation pressure is observed for subcooled water jets with high degree of subcooling, which is because the critical mass flux is overestimated using Homogeneous Equilibrium Model (HEM). For subcooled water jets with low degree of subcooling, HEM underestimates the data. Different from HEM, Henry-Fauske model always overpredicts the experimental data, where the degree of overestimation increases as the degree of subcooling decreases. The critical mass flux for saturated steam jets and saturated water/two-phase jets can be predicted well by isentropic expansion model and HEM, respectively.

Prognostics and Health Management (PHM): Where are we and where do we (need to) go in theory and practice

We are performing the digital transition of industry, living the 4th industrial revolution, building a new World in which the digital, physical and human dimensions are interrelated in complex socio-cyber-physical systems. For the sustainability of these transformations, knowledge, information and data must be integrated within model-based and data-driven approaches of Prognostics and Health Management (PHM) for the assessment and prediction of structures, systems and components (SSCs) evolutions and process behaviors, so as to allow anticipating failures and avoiding accidents, thus, aiming at improved safe and reliable design, operation and maintenance.There is already a plethora of methods available for many potential applications and more are being developed: yet, there are still a number of critical problems which impede full deployment of PHM and its benefits in practice. In this respect, this paper does not aim at providing a survey of existing works for an introduction to PHM nor at providing new tools or methods for its further development; rather, it aims at pointing out main challenges and directions of advancements, for full deployment of condition-based and predictive maintenance in practice.