Typical Nuclear Power Plant Research Articles

Online and Offline Fault Detection and Diagnostics in a Nuclear Power Plant Condenser

Nuclear power plants (NPPs) face significant financial pressures due to operational and maintenance costs. This research investigates Fault Detection and Diagnostics (FDD) techniques to optimize maintenance schedules and reduce expenses. The NPP condenser plays a critical role in converting turbine exhaust steam back into water for reuse. Condenser tube fouling, a prevalent fault mode, impedes heat transfer efficiency and can lead to decreased plant efficiency and safety risks. This study proposes an FDD framework that leverages raw signal analyses from temperature and pressure monitoring to detect and diagnose condenser tube fouling in both online and offline settings. The online approach facilitates close-to-real-time predictions, enabling proactive maintenance strategies. Additionally, the framework explores incorporating a condenser’s maintenance history for enhanced diagnostics. We employ a dataset obtained from a simulated nuclear power plant condenser using the Asherah Nuclear Power Plant Simulator (ANS). ANS replicates the operational dynamics of a pressurized water reactor (PWR) type NPP. The proposed methodology utilizes an encoder-decoder (E-D) structured 1DCNN model to analyze the raw signals. The research demonstrates consistent and accurate fault detection and diagnostics for condenser tube fouling in both online and offline scenarios. A high potential for generalization to unseen conditions was observed. However, online detection using small data windows necessitates caution due to potential false alarms around the transition points. Our findings pave the way for further exploration of robust diagnostics by accommodating a wider spectrum of fouling rates within degradation classes using ANS. This combined online and offline FDD approach offers a promising solution for promoting operational safety, efficiency, and cost-effectiveness in NPP condensers.

External Hazard Analysis for Water-Based Nuclear Power Plants

Based on recent studies, Water-Based Nuclear Power Plants (NPPs) include onshore and offshore floating NPPs, gravity-based structure NPPs, and underwater NPPs. These NPPs are increasingly considered for their benefits, such as the floating NPPs ability to distribute energy to remote areas, offshore sites, and construction activities. Regarding safety, external hazards must be considered during the design phase as stated in the Head of Nuclear Energy Regulatory Agency Regulation Number 3 Year 2011 and Specific Safety Requirements-2/1 (Rev 1). The external hazards of Water- Based NPPs should differ from those of land-based NPPs. Several studies have outlined the external hazards that water-based NPPs only address, such as onshore floating NPPs, and have yet to discuss the impact on water-based NPPs' safety. This study reviews the external hazards that must be considered and their effects on all types of Water-Based NPPs. The study is conducted by reviewing national regulations and international references to identify specific external hazards for Water-Based NPPs, followed by examining scientific works related to marine safety and Water-Based NPP design information to understand the impact of external hazards on the facilities. The review indicates that the external hazards that need to be considered for Water-Based NPPs include weather hazards, human- induced external hazards, earthquakes, and tsunamis. Keywords: NPP, Floating NPP, External Hazard, nuclear safety

Time-Series Forecasting of a Typical PWR Undergoing Large Break LOCA

In this work, a machine learning (ML) metamodel is developed for the time-series forecasting of a typical nuclear power plant response undergoing a loss of coolant accident (LOCA). The plant model of choice is based on the APR1400 nuclear reactor. The key systems and components of APR1400 relevant to the investigated scenario are modelled using the thermal-hydraulic code, RELAP5/MOD3.4, following the description published in the design control document. The model is tested under a spectrum of initial and boundary conditions via propagation of key uncertain parameters (UPs) which are derived from the phenomena identification and ranking table (PIRT). This is achieved by loosely coupling RELAP5/MOD3.4 with the statistical tool, Dakota. The most probable nuclear power plant (NPP) response was calculated using the best estimate plus uncertainty (BEPU) approach. Next, the database generated from the NPP system response was used as an input for the ML model. The NPP system response was represented by peak cladding temperature (PCT), safety injection system (SIT), mass flow rate, reactor power, and primary system pressure. In this research, two regression models were tested with reasonably good performance, namely, the gated recurrent unit (GRU) and the long short-term memory (LSTM).

Quantification of the Effective Detectable Period for Concrete Voids of CLP by Lock-In Thermography

This study is to inspect the voids between the concrete containment building and the containment liner plate (CLP) in the light-water reactor type nuclear power plant with lock-in thermography (LIT) inspection technology. For that, a finite element method (FEM) model containing concrete voids was created, and the thermal distribution change of the CLP surface was simulated through numerical analysis simulation of various LIT inspection conditions and converted with real-time thermography data. For the simulated temperature distribution image and the amplitude and phase images calculated by the four-point method, the signal-to-noise ratio (SNR) is analyzed based on the sound area and void areas. As a result, the difference in SNR according to the size of voids was remarkable, and the effective detectable period (EDP), which was common to each inspection condition, was derived. Furthermore, a CLP concrete mockup identical to the model shape is produced, and the thermal image of the EDP is analyzed through the experiment with the same analysis technique, and the results are compared. Although there are some differences between the numerical analysis conditions and the experimental environments, the deduction and utilization of EDP through FEM simulation are considered useful approaches to applying LIT to inspect concrete voids on the back of the CLP.

Interactive iterative methodology for seismic analysis of pile-raft supported nuclear power plant structure in a conventional framework

Due to the fast growth of the nuclear-powered industry, nuclear power plants (NPP) are being planned on soil sites, especially in active seismic zones. In such cases, a combined pile-raft foundation (CPRF) is considered a preferred choice for designers to support massive and stiff nuclear structures. As such, no study has been stated on the seismic design approach for NPP supported on the CPRF system. In the present study, an interactive iterative practical seismic design approach is proposed to estimate the dynamic response of a NPP resting on the CPRF considering continuum-based geotechnical and spring-supported structural models. The pseudo-static response of the continuum-based model is used to determine the soil-pile system stiffness for subsequent seismic analysis in a spring-supported structural model. A step-by-step algorithm is proposed to arrive at the final converged state of the combined system through an iterative procedure. A reasonably good agreement is achieved between the outcomes obtained from the suggested approach and the results of the experimental testing. The proposed approach is subsequently implemented on a typical NPP structure, and a good match is obtained with the results of the coupled time domain direct analysis, thus showing the reliability and practical applicability of the suggested approach. The suggested procedure can be used for the seismic analysis of NPP structures supported by the CPRF system.

Descriptor sliding mode observer based fault tolerant control for nuclear power plant with actuator and sensor faults

In sophisticated and complex system such as nuclear power plant, fault estimation and fault tolerant control always play an important role in maintaining the system stability and assuring satisfactory and safe operation. Thus, in this work a fault estimation and fault tolerant control scheme based on sliding mode theory is proposed for a pressurized water reactor type nuclear power plant considering simultaneous actuator and sensor faults. First, using descriptor sliding mode observer approach, an accurate estimation of the system states and sensor fault vector have been obtained simultaneously. Then, based on the estimated information, an integral type sliding mode control scheme is proposed to stabilize the resulting faulty system. With the help of Lyapunov stability theory, reachabilities of the proposed sliding mode surfaces are shown in both the state estimation space and the error estimation space, simultaneously. Finally, the efficacy of the proposed control scheme is shown by applying it to a nuclear power plant.

Assessment of tritium monitoring for radiation environment around the typical nuclear power plants in China

Assessment of tritium monitoring for radiation environment around the typical nuclear power plants in China

Using machine learning to forecast and assess the uncertainty in the response of a typical PWR undergoing a steam generator tube rupture accident

In this work, a multivariate time-series machine learning meta-model is developed to predict the transient response of a typical nuclear power plant (NPP) undergoing a steam generator tube rupture (SGTR). The model employs Recurrent Neural Networks (RNNs), including the Long Short-Term Memory (LSTM), Gated Recurrent Unit (GRU), and a hybrid CNN-LSTM model. To address the uncertainty inherent in such predictions, a Bayesian Neural Network (BNN) was implemented. The models were trained using a database generated by the Best Estimate Plus Uncertainty (BEPU) methodology; coupling the thermal hydraulics code, RELAP5/SCDAP/MOD3.4 to the statistical tool, DAKOTA, to predict the variation in system response under various operational and phenomenological uncertainties. The RNN models successfully captures the underlying characteristics of the data with reasonable accuracy, and the BNN-LSTM approach offers an additional layer of insight into the level of uncertainty associated with the predictions. The results demonstrate that LSTM outperforms GRU, while the hybrid CNN-LSTM model is computationally the most efficient. This study aims to gain a better understanding of the capabilities and limitations of machine learning models in the context of nuclear safety. By expanding the application of ML models to more severe accident scenarios, where operators are under extreme stress and prone to errors, ML models can provide valuable support and act as expert systems to assist in decision-making while minimizing the chances of human error.

Key elements for proper development and implementation of severe accident management guidance/guidelines – SAMG

Severe Accident Management Guidelines (SAMG) have been developed to support a nuclear power plant (NPP) in case of a severe accident, i.e. an (unlikely) accident where the core and/or the fuel in the Spent Fuel Pool (SFP) have been damaged and there is risk for radioactive releases. Various approaches exist, such as the SAMG developed by the PWR and BWR Owners Groups (PWROG, BWROG), by Electricité de France (EdF) for its fleet of plants, the Candu Owners Group (COG) and by several individual NPPs. Often, the developers of SAMG develop a generic type of SAMG, applicable for a certain type of NPP, and leave it to the individual plants to adapt these to their configuration and Emergency Response Organisation (ERO). That means, that quite some burden remains for the individual plants to finally implement the SAMG. This implementation includes not only technical items, such as defining proper strategies, but also many organisational issues, as handling a severe accident needs an effective ERO, in which the SAMG must be well embedded.Various plants have been reviewed, both the ways they have constructed their plant-specific SAMG, implemented them in their ERO and tested the performance in drills and exercises. Most recently, a review of SAMG at Swedish plants has been performed, after which it was decided to pull the various experiences together and summarize them in one compendium. Some of these also include aspects of the development of the generic SAMG.The result is a number of recommendations (∼30), both on the technical field as well as regarding the ERO. We could also define a number of good practices that deserve wider attention.A few words have been devoted to international support for a stricken plant (Epilogue).

Cooling water intake system safety analysis based on impingement probability

Invasion or aggregation of marine organisms in cooling water intake systems (CWIS) has gradually become an important problem affecting the safety of nuclear power plants with environmental and climate changes. In this study, a 3-dimensional numerical model (TELEMAC-3D) was used to determine the impingement probability in a typical nuclear power plant with a once-through cooling system, and the effect on CWIS safety. The factors controlling impingement probability were also analyzed. Results show that (1) impingement probability decreased rapidly with an increase in distance from the CWIS. In addition, the distance of the impingement effect of a nuclear power plant with six units was mainly within 1 km of the CWIS. (2) Impingement probability increased with water withdrawal, and as distance to the CWIS increased, the increase in probability increased. (3) Generally, an increase in tide strength led to a decrease impingement probability. (4) Near the CWIS, the impingement probabilities of areas upstream or downstream of the CWIS along the tidal flow direction were much higher than those not in those areas. (5) An increase in water depth significantly reduced impingement probability. When the water depth of the CWIS increased from 5 m to 15 m, impingement probability was reduced up to 30%. Based on the above findings, the following suggestions were made to minimize the impingement effects on CWIS safety: first, the CWIS of coastal nuclear power plants should be set in an area with low aquatic biomass, strong tides, deep water, and few surface species within the range of 1 km, and second, the amount of cooling water withdrawal or velocity should be reduced as much as possible.

Multi-criteria decision model for selection of nuclear power plant type

Interests in clean energy revived the nuclear power industry. For the first time in decades, new technologies and plant designs are being considered by regulatory agencies. This shift in both public opinion and regulatory requirements opens opportunities for decision-makers and calls for a decision-making tool to help determine a feasible nuclear power plant (NPP) type. This paper contributes to academic literature by presenting an Analytic Hierarchy Process model to address the problem of selecting a feasible NPP type for construction. A holistic approach is taken, considering high level criteria in the categories of economic, operational, and risk concerns. Subject Matter Expert inputs to the pairwise comparison of priorities and criteria selection aided the analysis. Results show that the traditional NPP designs are closely followed in preference by a newer Small Modular Reactor (SMR) design. From this analysis, several areas of expansion and future work were identified, in hopes of growing the decision-making studies in NPP selection and including more holistic approaches when considering the best NPP type for new construction.

Small modular reactors: An assessment of workforce requirements and operating costs

In the past seven years, global warming discourse has revolved around restricting temperature increases to under 2 °C above pre-industrial levels by utilizing such low-carbon energy sources as nuclear, solar, and wind energy. However, the use of nuclear energy for electricity generation via conventional nuclear power plants (NPPs) has met stiff challenges due to high capital costs, thereby constraining newcomer nations. A solution contending high capital costs for NPPs is the introduction of small modular reactors (SMRs) that has garnered attention over the past ten years because of their lower initial capital investment, scalability, and ease of deployment for less developed electricity grid and small areas of low electricity demand. In this study, the workforce requirements and salaries component for operating workers of 300 MWe integral Pressurized Water Reactors (iPWR) SMR is modeled and assessed using an upgraded International Atomic Energy Agency (IAEA) Nuclear Power Human Resources (NPHR) modeling tool from 2018 to 2055. The results show that at a discount rate of 10% an average annual salary payment, the net present cost (NPC) for the operating workers is approximately 431 million USD with a standard deviation of 8%. Moreover, the results further show that at the end of themodeling period and considering 43 function areas in a typical NPP, the top three functions to be recruited and contributing to the total domestic workers for a typical newcomer nuclear nation are Maintenance/Construction (18.17%), Operations (9.02%), and Management (8.22%) workers. The results enable the estimation of human resource requirements to operate an SMR over its life cycle, which in turn enables informed budgeting for human resources.

Role of welding residual stress in stainless steel piping with application of the leak-before-break technology

The predicted leakage rate of piping circumferential through-wall cracks (CTWC) under various loading levels is a critical factor for the application of leak-before-break (LBB) technology. In current engineering approaches, the effect of welding residual stress has not been carefully taken into account. In this paper, both numerical analyses and comparative verification are adopted to examine the influence of typical welding residual stress field on the crack opening displacement for austenitic stainless steel piping with representative geometric dimensions and in situ measured material performance curves. An in-depth investigation is carried out to reveal the effect of the residual stress on morphological parameters of the CTWC flow channel. In the case where the residual internal stress of welding has the greatest effect, the Henry-Fauske model is employed to analyze the flow medium passing through the CTWC and the corresponding leak rate for a typical nuclear power plant. The results indicate that the welding residual stress leads to a substantial change in the crack opening displacement and crack morphology parameters. Both the current GE/EPRI method and the NUREG/CR-6837 modification recommended by the American Electric Power Research Institute underestimate the effect of this phenomenon, resulting in a higher prediction of the medium leakage rate. A similar situation is most likely to occur for the cases of short cracks in thin-walled piping and long cracks in thick-walled piping. Additionally, the obtained results reveal that the welding residual stress causes the whole crack surface to open negatively, and the axial line of the long crack on the thick-walled piping is close to a conic curve under specific conditions.

Comparison of Estimates of Costs involved between two Decommissioning Strategies for PWR type Nuclear Power Plants

This work compares the project costs of two possible decommissioning strategies for a hypothetical nuclear site with similar characteristics to the CNAAA reactors. The cost of the process is sensitive to the type of strategy adopted, as the type of strategy involves the duration of the process. Defining project strategy/duration and estimating cost are part of a coupled problem. Thus, decommissioning cost is a relevant step and needs to be rigorously defined before starting the decommissioning process, as demobilizing teams/equipment, and remobilizing entail extra costs. Cost estimates follow the Av-Descom model. Tool easily implemented by a spreadsheet-like code. Among the stages of the two strategies presented, the one with the greatest impact on the cost of the project is the stage referring to the transition period comprised by the decontamination tasks of the systems, equipment, and structure of the deactivated NPP. This step exposes workers to radiation doses and to the greatest occupational hazards. It was observed that the cost variation from one strategy to another is above 50% of the total cost of the project. It was concluded that defining the best decommissioning strategy is complex and important to be defined from the beginning of the decommissioning process. It was inferred that the strategy that best fits the Brazilian reality is the delayed decommissioning.

Evaluation of seismic margins for nuclear power plant piping systems using ratcheting criterion

In the wake of recent beyond design basis earthquakes, it is very essential to evaluate seismic margins of nuclear power plant piping systems. Better seismic margins will ensure the safety in the event of less probable bigger seismic event beyond consideration of design level. At present, permissible codal stress limits are used to evaluate these margins and it has been experimentally demonstrated that ratcheting is the significant failure mode for piping systems subjected to reversing dynamic load. Hence, it is more prudent to use actual failure criterion for evaluation of seismic margins. In this paper, a detailed framework is provided for evaluation of seismic margins of piping systems using ratcheting criterion. As ratcheting is the potential mode of failure for piping systems under earthquake load, two strain-based performance levels based on code practice and experience gained with experiments carried out by Bhabha Atomic Research Centre, are used for margin assessment. It is usual practice to carry out stress-based response spectrum analysis for design and qualification of piping systems under earthquake load. In order to offer a practical, simpler and feasible solution for ratcheting evaluation, the same response spectrum analysis in iterative way, can be used to evaluate ratcheting in the piping systems. Experimentally validated numerical procedure is used for ratcheting evaluation. This procedure is demonstrated for a piping system of a typical nuclear power plant. Seismic margins for the piping system are evaluated using codal provision and two ratcheting based performance levels for deterministic and probabilistic methods.

Methodology for Online Sensor Recalibration

A significant contributor to nuclear power plant operations and maintenance (O&M) costs is the periodic calibration check of sensors. Periodic calibration checks provide the necessary confidence that the measurements from these sensors are correct, with the data used to monitor and verify proper operation of the reactor. The periodicity of calibrations in the nuclear industry can range from once in several weeks for some instrument channels to once every refueling outage (~18 months) for certain safety-significant pressure and level transmitters. With expected longer refueling intervals in many advanced reactor/small modular reactor concepts, fewer opportunities are expected to be available for manual calibration checks and recalibration for many instrument channels.
 Presently, periodic sensor calibration checks are performed manually, with manual recalibration performed if the sensor is found to be out of calibration. Although studies have shown that most (over 90%) sensors are found to stay within calibration specifications over a calibration cycle (~18 months), labor must still be spent to verify that these sensors are within calibration. Given the large number of sensors (anywhere between 100–2400 sensors) in a typical nuclear power plant, the ability to identify sensors that are failing/failed or drifting out of calibration and limit recalibration to those specific sensors has the potential to save between $0.5–1M per year per plant. The cost of calibration checks and recalibration are expected to be of greater significance to advanced reactors and small modular reactors, given the industry focus on reducing O&M costs for these reactor concepts as part of improving the economic viability of advanced nuclear power.
 Methods that can compensate for calibration drift by adjusting the calibration automatically and in real-time may be able to further reduce costs associated with recalibration. Such autocalibrations or online automated recalibrations reduce unavailability of instrumentation and improve online maintenance planning flexibility from both resource allocation and online risk evaluation perspectives.
 This paper describes an initial set of algorithms developed for the purpose of detecting drift and correcting for it through an online recalibration method. Initial results on laboratory-scale experimental data indicate the potential of these algorithms to detect calibration drift and update calibrations with prediction and drift correction accuracy exceeding 95%.

Assessment and design of nuclear power plant structures under external explosions

With the increasing concerns over the safety of the nuclear power plants (NPP), the blast-resistant assessment and design of NPP structures have been emphasized both domestically and internationally. Aiming to efficiently assess and design the NPP structures under external explosions, the flexural SDOF method and the checking method of direct shear failure were firstly established, and the critical parameters concerning the blast loadings and structural resistance function, were determined. Then, a general assessment and design method for blast-loaded RC structures was proposed, and validated by comparisons with the experimental maximum lateral deflections and whether a direct shear failure occurs. Finally, a practical response surface method (RSM)-based design formula was proposed for the typical NPP structures. It is found that, (i) the pressure-based SDOF method has a higher prediction accuracy compared with the velocity-based SDOF method; (ii) with the increase of the slab thickness and rebar diameter, both the values of reflected pressure and impulse asymptotes increase; (iii) the typical NPP structure remains intact under blast loads specified by the Chinese Load Code for the Design of Nuclear Power Plants Building Structures; (iv) increasing the steel and concrete strength can enhance the blast resistance of the NPP structure at the plastic state attributed to the increased ultimate structure resistance, but not at the elastic state; (v) the design charts for typical NPP structures under the blast loads regulated in Safety Reports Series No. 87, Safety Aspects of Nuclear Power Plants in Human Induced External Events: Assessment of Structures, are given.

Graded Approach Establishment for the HTGR Maintenance Activities Using Modified Fuzzy FMEA & Expert Judgement Methodology

Grading is an important step of the Nuclear Power Plant (NPP) operation & maintenance activities. However, there are several grading difficulties for the High Temperature Gas Reactor (HTGR) as well as the other type of NPPs causing by the lack of operational experiences and availability of the reliability data. Failure Mode & Effects Analysis (FMEA) is one of the mature techniques that are commonly used to solve such kinds of difficulties. Nevertheless, traditional FMEA has several issues and possibly become an obstacle in the grading process. The modified FMEA by utilizing expert judgment elicitation techniques combined with the fuzzy logic theory is proposed to solve those issues. As a study practice, the proposed methodology is applied by examining Japanese’s HTGR, Gas Turbine High Temperature Reactor 300 for Cogeneration (GTHTR300C) design carefully. This study establishing good practice especially for the future advanced NPP maintenance activities development.

Studi Karakteristik Termohidrolik Pada Kanal Pendingin Teras Reaktor SMART Menggunakan CFD Ansys Fluent

The Small Modular Reactor (SMR) type nuclear power plant is currently growing very rapidly by offering various kinds of safety features, both active and passive. One of the SMRs currently to be built and operated is the SMART (System-integrated Modular Advanced Reactor) reactor. The SMART reactor is designed to use conventional PWR fuel types but operates at a lower flow rate and core inlet temperature than conventional PWR. The existence of this basic difference needs to be a separate note for us in evaluating the safety of the SMART Reactor, especially for the thermohydraulic aspect. The thermohydraulic characteristics have been carried out previously using the MATRA code, however, the temperature distribution in the cooling sub channel is still unknown. Therefore, the purpose of this study is to calculate the temperature distribution in the cooling sub-channel on SMART. The program used in this study is CFD ANSYS. Ansys Fluent 2021 R1 Computational Fluid Dynamic (CFD) computer code was used to simulate coolant flow and heat transfer from fuel to coolant. The cooling sub channel was modeled consisting of four fuels with a diameter of 0.95 cm and a pitch of 1.26 cm in a rectangular arrangement surrounded by light water coolant. Based on the calculation, there is an increase in the average temperature of the coolant between the input and output sides of the coolant sub channel of 60 ℃, which is greater than the SMART reactor design. The fuel center and fuel cladding temperatures were calculated at 1100 ℃ and 350 ℃, respectively. This difference is caused by the modeling assumptions which is still far from the actual design. The model was assumed to consist of pure fuel and ignores the actual variations in fuel and fuel assembly. Keywords: SMART Reactor, CFD, Ansys Fluent, Thermal Hydraulic

Minimum dose path planning during reactor coolant system maintenance with corrosion product activity

To facilitate the periodic maintenance of the reactor coolant system, the nuclear power plant must be shut down. However, maintenance work on the primary loop components could result in post-shutdown radiation exposure. Activated corrosion products are the main source of radiation inside the containment building during post-shutdown. Traditionally, radiation detectors are used to measure in-containment radiation to protect workers from radiation risk during maintenance. However, these devices cannot predict corrosion product activity and the subsequent radiation dose. This study presents a minimum dose path planning methodology for protecting workers during the reactor coolant system maintenance in a typical pressurized water reactor nuclear power plant. Specifically, the study uses the corrosion product release and activation (CoPRA) code to estimate the corrosion product activity from the primary loop’s three major post-shutdown source terms. Subsequently, the FLUKA Monte Carlo code is used to simulate dose rates from the corrosion product activity, and the Three-Degree of Vertex shortest path algorithm is implemented to find the minimum dose path. Comparison results show that the predicted dose is consistent with the measured dose from the nuclear power plant. The results also show the flexibility and adaptability of the proposed method in finding the shortest path with the minimum dose in different reactor operating conditions.