Abhinav Gupta

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The Center for Nuclear Energy Facilities & Structures, has been established and is administered by North Carolina State University to conduct research in the areas of structural engineering, mechanics, risk assessment, hazard mitigation, and construction engineering and to promote research, education, and training in the Research Area. The CENTER has developed core research, non-core research, and technology transfer activities.

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The proposed project builds upon the previous work of CNEFS in which CNEFS helped develop capabilities for Fault Tree Analysis in the Idaho National Lab (INL)������������������s MASTADON toolkit. The proposed project focuses on the following specific tasks ��������������� Extend the fault-tree analysis and quantification to include event-trees ��������������� Convert C++ code to MOOSE objects and implement in MASTODON. ��������������� Create examples for PRA in MASTODON using the new fault-tree and event-tree quantification implementations ��������������� Benchmark examples with Saphire ��������������� Document these examples on the MASTODON website

In recent years, there is an increasing interest in the nuclear industry to focus on identifying tools, methods and opportunities to optimize construction activities and reduce costs of operation and maintenace. One of the promising tools is the use of digital twins. A digital twin is a continuously updated representation of an actual structure as it degrades. It uses the observations from maintenance and sensor data as input to continuously update the simulation and data-driven models while considering the effect of uncertainties. There is a need for more demonstrations of digital twins use cases to open the door for more nuclear industry applications. Conduct an exploratory project to demonstrate the various steps needed in the development of a digital twin on a piping system and to develop a computational framework for assessing degradation mechanisms. To achieve the high level objective, the contractor will build a piping system consists of individual pieces of pipes, elbows and flanges. The details of the piping system will be discussed with the EPRI project manager (PM).

The main objective of the proposed work is to develop, demonstrate, and evaluate a probabilistic risk assessment (PRA) software platform needed to address the major challenges of the current legacy PRA tools, such as better quantification speed, integration of multi-hazard models into traditional PRAs, and model modification simplification and documentation automation. To achieve the main objective, we will first perform benchmarking and profiling of current PRA tools, such as SCRAM and SAPHIRE, to investigate the current bottlenecks in the quantification speed and memory requirements. Secondly, we will design, implement, and benchmark a PRA software platform based on a web-based stack using the latest technologies available to overcome the mentioned challenges. Finally, we will evaluate the performance gains of this framework by modeling and quantifying large PRA models that would have been too expensive to run using the legacy PRA tools.

A DT-DAP (Digital Twin Development and Assessment Process) methodology has been formulated at NCSU in the ARPA-E sponsored project. DT-DAP can be very effective in guiding the design, training, testing, and application of DTs to improve effectiveness, accuracy and acceptance of system design and safety analysis.

Probabilistic seismic hazard analysis (PSHA) is a key component in seismic design and performance assessment of engineering components. Accurate representations of rupture characteristics, wave propagation, and subsurface soil behavior are necessary to perform an accurate PSHA. However, in traditional PSHA, simplified empirical Ground Motion Models (GMMs) are used to estimate the ground motion levels. These GMMs neglect the inherent physical complexities in earthquake rupture and ground motion properties, such as slip heterogeneity, rupture directivity, and basin depth, among others. In addition, the paucity of ground motions recorded from large magnitude ruptures in the near-fault region, and from stable continentals regions (e.g. French Context) makes GMMs unsuitable for several contexts and applications. The objective of this research is to: 1) apply seismic ground motion methods to generate synthetic ground motions data set based on a real ground motion data set, 2) make use of synthetic ground motions to update GMMs using the Bayesian method already developed in a precedent collaboration (NCSU-EDF) and/or to develop a specific synthetic GMM (using only simulated motions) and/or to develop a hybrid GMM (using synthetic and real ground motions). The challenge of this research is to propose new alternatives to the PSHA based on simulated ground motions to complete the logic tree branches. The weight of logic tree branches of different simulated approaches and GMMs will be assigned to the validity of simulated and empirical models, with respect to the observed ground motions in regions under evaluation.

Over the past decade, the use of artificial intelligence techniques in the field of health-monitoring has gained significant interest, especially for structures such as building and bridges. This project proposes development of an Artificial Intelligence (AI) framework for the data-driven condition monitoring of nuclear structural systems and equipment, where the vibration response is governed by multiple localized modes unlike that in buildings and bridges. Hence, techniques such as signal processing and pattern recognition will be employed to extract degradation-sensitive features. Degraded locations can potentially exhibit damage such as localized yielding, cyclic fatigue, or initiation of cracking. Moreover, such locations can at times go undetected by current inspection techniques. Therefore, this research proposes a framework, which utilizes sensor response to generate an AI database for predicting degraded locations and severity in nuclear structural systems and equipment. Degradation severity will be classified as minor, moderate, and severe, along with incorporation of uncertainty.

There have been and will continue to be rapid advances in 3D scanning and augmented/virtual reality technologies to improve construction costs and schedules, especially in the nuclear energy industry that has suffered from construction cost escalation and delays. A key challenge faced in the implementation of these modern technologies relates to changes needed in the regulatory practice and approvals. It is vital for regulatory agencies like USNRC to understand the fundamental basis of these technologies and characterize the accuracy and consistency in them. This project is aimed at conducting the research needed to support this effort which would eventually be needed in the near future. More specifically, it is proposed to: (i) characterize technical specifications and the associated parameters that govern the accuracy of virtual inspections, and (ii) identify items at a construction or manufacturing site that are conducive to virtual inspections and the items that are not.

The objective of the proposed research is to use advanced modeling and simulation tools to determine if the building-equipment interaction help in reduction of response of secondary systems when subjected to high frequency motions. The motivation for conducting the proposed research is driven by the anticipated savings in the enormous effort and cost that is currently faced by the nuclear industry in attempting to qualify equipment, piping, and structures for the updated seismic hazard containing high frequency motions.

In recent years, flooding at nuclear power plants (NPP) has increased emphasis on using high fidelity simulations to evaluate the vulnerability of nuclear plants. One of the key limitations in the use of high fidelity simulations is related to a lack of verification and validation (V&V) of such simulations. One outcome of incomplete and insufficient V&V relates to a large degree of uncertainty in the simulation results which in turn leads to conservative assumptions by the decision makers. Past experience has shown that such conservative assumptions in the context of safety assessment for other external hazards such as seismic have resulted in highly overdesigned NPP systems and excessively high costs. Therefore, it is quite important that various uncertainties in this process are appropriately identified and included through formal uncertainty quantification (UQ). A robust framework for verification and validation is needed to not only include uncertainties but also to formalize the confidence in predictions of system level validation that are based on component level data using Bayesian Network. In addition, the framework has to identify the basic events that are critical in the perspective of overall validation. This process helps in allocating the resources efficiently thereby reducing the effort to conduct high fidelity simulations and large-scale experiments.

Huge escalations in overnight construction costs and schedule delays have rendered nuclear energy commercially unattractive. Much of the research and development has focused on developing new reactor designs with accident tolerant fuels and passive safety systems intended to reduce operating and lifecycle costs. There has been little to no investment in research on nuclear construction management. Moreover, Hopf [1] argues that the primary reasons for construction cost escalation and schedule delays are related to: (i) extremely stringent nuclear safety and quality assurance standards, (ii) inexperience in managing and staffing to nuclear QA standards, (iii) excessive paperwork, and (iv) supply chain delays due to rework. To ensure the success of implementation and deployment of a new reactor (i.e., Versatile Test Reactor (VTR)), design, construction, and testing processes should be integrated. Virtual Design and Construction (VDC) using Building Information Modeling (BIM) has been proven to be a solution that bridges gap between design and construction [cite]. Furthermore, it has been used as a project control tool [cite]. BIM as a central hub of digital data exchange can bring different experts (i.e., engineers with different disciplines and contractors who build and manage construction projects) virtually together. All stakeholders have a shared resources of knowledge to help identify and minimize risk during design, construction, and operation and maintenance phases [cite]. This proposal proposes to investigate and help INL better understand the capabilities of BIM that can support the integration of design, construction, and testing of the VTR. Furthermore, the research team will pick one of the analysis (i.e., pipe stress) and develop a software/plug-in that produces 2D/3D models that can be directed imported by the analysis tool.

In recent years, flooding at nuclear power plants (NPP) has increased emphasis on using high fidelity simulations to evaluate the vulnerability of nuclear plants. One of the key limitations in the use of high fidelity simulations is related to a lack of verification and validation (V&V) of such simulations. One outcome of incomplete and insufficient V&V relates to a large degree of uncertainty in the simulation results which in turn leads to conservative assumptions by the decision makers. Past experience has shown that such conservative assumptions in the context of safety assessment for other external hazards such as seismic have resulted in highly overdesigned NPP systems and excessively high costs. Therefore, it is quite important that various uncertainties in this process are appropriately identified and included through formal uncertainty quantification (UQ). A robust framework for verification and validation is needed to not only include uncertainties but also to formalize the confidence in predictions of system level validation that are based on component level data using Bayesian Network. In addition, the framework has to identify the basic events that are critical in the perspective of overall validation. This process helps in allocating the resources efficiently thereby reducing the effort to conduct high fidelity simulations and large-scale experiments.

The proposed project seeks to establish a technical basis for, and preliminary development of, a Nearly Autonomous Management and Control (NAMAC) system in advanced reactors. The system is intended to provide recommendations to operators during all modes of plant operation except shutdown operations: plant evolutions ranging from normal operation to accident management. These recommendations are to be derived within a modern, artificial-intelligence (AI) guided system, making use of continuous extensive monitoring of plant status, knowledge of current component status, and plant parameter trends; the system will continuously predict near-term evolution of the plant state, and recommend a course of action to plant personnel.

This project will develop methodology for verification and validation of advanced computer models used in Risk-Informed Safety Margin Characterization (RISMC) of nuclear power plants. The project will apply the methodology to selected problems in nuclear reactor safety. The focus is placed on computer codes that simulate external hazards that have impact on plant safety, including severe accident management and emergency response. The project will collect, characterize, archive, and use data from plant measurements, integral-effect and separate-effect tests to support code validation. The methodology will bring to use probabilistic risk assessment (PRA) to guide a risk-informed validation approach implementation.

The objective of the proposed research is to use advanced modeling and simulation tools to determine if the nonlinearities in the mounting arrangement of the cabinets or building equipment interaction filter out the high frequency motions. The motivation for conducting the proposed research is driven by the anticipated savings in the enormous effort and cost that is currently faced by the nuclear industry in attempting to qualify vibration sensitive digital and electro-mechanical devices for the updated seismic hazard containing high frequency motions.

Traditional concrete protective structures encountered by ground forces typically have unconfined compressive strengths of between 3,000 to 6,000 psi. Recent advances in concrete technology have resulted in new concrete materials with compressive strengths of 30,000 psi or greater. No field instrument currently exists that can simultaneously determine the compressive strength, thickness, and rebar configuration of concrete structures across today������������������s wide range of possible compressive strengths. In STTR Phase I/II efforts, NLA Diagnostics LLC (NLAD) developed a prototype instrument that measures both compressive strength and concrete thickness up to 33,000 psi and 6 ft, respectively. NLAD has also mapped out a technical course to locate rebar within the concrete. This proposed project will take input from military operators to improve packaging of the existing technology and to implement a rebar mapping capability in order to enhance military capabilities to defeat concrete protective structures, accelerate the ability to deploy concrete characterization instruments with operating forces, and reduce the technical risks associated with engaging concrete protective structures in tactical situations.

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Leakages in piping systems or their connections to the vessels and pumps can result in small LOCA which has the potential to go undetected. A seismic event can either directly initiate such a leakage or cause enough damage that would eventually lead to such a leakage during post seismic operation. The state-of-practice in the nuclear power industry relies solely on a linear seismic analysis of piping systems which has many limitations particularly with respect to the capability for evaluating and detecting any possible risk of leakage and corresponding locations. The key limitations can be summarized as: ��������������� Lack of commercial computer codes to incorporate nonlinearity at localized locations of joints or nozzle connections in a piping analysis prohibits consideration of aging in the simulation models. ��������������� Commercial piping analysis codes consider piping systems to be uncoupled from the supporting building and equipment. This uncoupling ignores the interaction between the building-equipment-piping systems which can be quite significant. It has been shown that appropriate consideration of such interaction gives piping stresses which are usually an order of magnitude lower than those calculated from a conventional uncoupled analysis. ��������������� Conventionally calculated excessively high forces and moments have led to an industry wide use of special seismic supports such as snubbers or gap supports. However, such supports make the piping systems stiffer and in turn they naturally attract greater forces during operational vibrations or during any seismic event.

It is proposed to implement the closed-form formulations for coupled seismic analysis of primary-secondary (building-equipment) systems into a spreadsheet and use them to conduct a suite of analyses for a series of representative systems as well as representative realistic systems by using modal properties (frequencies, participation factors, and mode shape ordinates at the desired degrees of freedom). The simple and representative systems will be developed in close coordination and consultation with the sponsor. Consider parameter variations and different earthquake motions to statistically evaluate factors for degree of reduction due to coupling. These factors can then be used more widely and generically and would be technically justified for use in SPRAs.

The project is to investigate capability of a computational fluid dynamics method, named Smoothed Particle Hydrodynamics, for simulation of high wind scenarios and high wind impact on nuclear energy facilities.

The Fukushima accident showed the need to explore scenarios where external hazards exceed the design basis. A risk-informed framework and fundamentally sound methodologies for fragility assessment can assist in evaluating the performance of a plant not only for design basis external hazard but also for vulnerabilities beyond the design basis. The assessment of an operating facility against extreme external hazards should therefore review the response of the installation to the hazards within the context of uncertainties in predicting hazard as well as the uncertainties in a plant������������������s SSCs to a particular hazard. The proposed approach in this research is intended to evaluate (a) the level of hazard that compromises the safety of a plant, (b) potential gaps in the defence in depth provisions, and (c) improve overall safety by improving the fragilities of individual components that contribute to the overall plant risk.

It is proposed to further build upon the closed formulations being implemented in another closely related project sponsored by EPRI. Dominion will provide FE models of two structures in SAP code and the FIRS for these structures. NCSU researchers will make parametric adjustments (e.g. soil stiffnesses, represented by soil springs) and develop ISRS at key locations in these two structures that have similar shapes / amplitudes as the ISRS from current QA verified calculations. Dominion will provide a suite of representative equipment weights, natural frequencies and damping values. NCSU researchers will perform analyses using modal data of these two structures to develop ISRS that include coupling effects.

This proposal is aimed at assisting ABB with technical expertise regarding seismic qualification of GIS equipment particularly in relation to their project for San Mateo, CA. Typically, the GIS unit is placed on a floor which is essentially at the ground level. Furthermore, a particular type of GIS unit has been qualified by ABB using high level shake table test in which the ground motion had a PGA of 1.0g which is fairly large value compared to typical design PGA values. However, in San Mateo, the floor is located at an elevation of 14 ft above the ground. Consequently, the floor accelerations are expected to be significantly influenced by the dynamic characteristics of the building housing the unit. Also, the elevated floor location is likely to produce peak accelerations much higher than 1.0g.

Overall objective is to study the seismic performance of non-structural ceiling systems in buildings such as fire suppression piping, suspended ceiling fixtures, HVAC ducts, and partitions. NCSU?s work is focused on computer modeling, optimization, and fragility evaluations for designing piping configuration needed to conduct experiments, Subtasks of this study focus on verification of theoretical formulations for seismic analysis of coupled building-piping systems as well as development of new formulations for improved verification with respect to the experimental results obtained by other participating organizations.

This project focuses on probabilistic safety assessment of nuclear power plant structures using Bayesian approach. The project tasks are focused on providing mathematical formulations for the structural fragility assessment using Bayesian methods. An attempt will be made to propose an appropriate method for choosing sampling points in complex models when using Bayesian methods and on the possibility of developing an index which shows the importance of statistical evidence. The methodology will be illustrated by application to simple examples. Finally, we will provide a discussion on the merits and challenges of this approach with respect to its applicability in structural fragility assessments.

The purpose of this study is to develop and implement a generally applicable seismic fragility methodology for structural components and systems. The study will develop an approach that provides a closer integration of the characterization of earthquake ground motions and the performance of critical facilities. As the recent experience at the Kashiwazaki-Kariwa nuclear power plant during last July?s earthquake demonstrated, there is considerable margin in the design of well-built nuclear power facilities. Our job as researchers is to develop methods that provide a ?best?, neither conservative nor unconservative, assessment of the expected performance of these facilities. This research will develop an approach that takes advantage of Kajima structural analysis capabilities and integrates an improved characterization of earthquake ground motions with a detailed reliability analysis of structure performance.

This study provides an experiment-based application for the blast/impact software investigation carried out at INL. Active research collaboration between INL and the Center for Nuclear Power Plant Structures, Equipment and Piping at NCSU is proposed. This collaboration will provide NCSU students and faculty access to the advanced finite element software available at INL so that they can model the structural performance of FRC. Simple experiments on FRC structural members will be conducted at NCSU to reconcile analytical results and verify existing experimental results. The purpose is to further the development of new FRC materials to withstand impact and blast loads.

The 19th International Conference on Structural Mechanics in Reactor Technology (SMiRT 19) will be held in Toronto, Canada, August 12-17, 2007. Sponsored by the International Association for Structural Mechanics in Reactor Technology (IASMiRT), it is the leading international conference on structural safety in nuclear power reactors and facilities; it has been held biennially in Asia, Europe and the Americas since 1971. North Carolina State University's Center for Nuclear Power Plant Structures, Equipment, and Piping, in collaboration with the Canadian Nuclear Society and NC State's Office of Professional Development, is organizing the conference.