Nuclear Power Units Research Articles

Matching Analysis of Technical Parameters and Safety Standards for Nuclear Replacement of Coal-Fired Units

In the context of the growing share of renewable energy and the impending decommission of a large number of coal-fired units, nuclear energy is the only green energy that can replace coal power for a stable, clean, efficient, and large-scale power supply. This article compares the differences between coal power and nuclear power in terms of thermal system, thermal cycle, turbine parameters, and safety. It discusses the possibility of replacing the boiler of a coal power plant with nuclear power, that is, replacing the boiler of a coal power plant with a nuclear reactor for generation/heating/cogeneration. For coal-fired units with similar capacity that do not use a reheat cycle (at or below high pressure) and nuclear power units (such as a high-temperature gas-cooled reactor), as well as coal-fired units with a reheat cycle (ultra-high pressure and above) and nuclear power units (such as pressurized water reactors), there are great differences in steam parameters. In terms of steam turbines, the size of nuclear power units is relatively larger, requiring additional dehumidification measures. In addition, the safety factors and management methods considered in the site selection, construction, and operation of nuclear power plants are more stringent and complex, and comprehensive analysis and evaluation are needed in aspects such as waste treatment and accident emergency response. Except for the relevant provisions of the American Society of Mechanical Engineers code for pressure vessels, nuclear power units are not compatible with coal-fired units in terms of safety standards. Therefore, considering comprehensively, it is believed that the scheme of nuclear power replacing coal-fired units for power generation/heating/cogeneration program is not feasible at present.

Optimal design of circulating water system in pressurized water reactor nuclear power plant

The optimization of the circulating water system of pressurized water reactor nuclear power plants can effectively improve the profit without affecting the primary circuit operation, which is of great significance in enhancing the market competitiveness of nuclear power plants. The circulating water system of the nuclear power plant is modeled, and the circulating water system optimization scheme is formed after the nuclear power unit is reformed from three aspects: condenser heat transfer area, condenser connection mode, and circulating water pump operation mode. The circulating water system optimization calculation is carried out for each scheme by using the minimum annual cost method. Based on the sensitivity analysis of the factors that affect the profits of nuclear power plants, the best circulating water system optimization scheme is finally given for new nuclear power plants and nuclear power plants in operation, which provides a reference for the construction of related nuclear power projects.

The concept of autonomous hydrogen energy complex: Adaptation of large nuclear power units to uneven energy consumption schedules

Evaluation of flexible operation of nuclear power plants (NPP) during load reduction hours shows inefficiency of the process with account for high capital investments in NPPs at relatively low fuel costs. The concept of producing hydrogen fuel at NPPs during the hours of low power demand in the energy grid will be an effective solution to the problem. Generating electricity through utilization of electrolysis hydrogen and oxygen production provides a possibility to significantly extend the flexibility range for power released by the NPP in the energy grid under a constant reactor capacity factor. Utilization of a low-power steam turbine can be a good solution to address the challenges of electricity production from hydrogen and oxygen. This will ensure high reliability of the main NPP steam turbine. In addition, in their previous works based on the experimental data relating the operation of NPPs, the authors showed that a low-power steam turbine and residual energy release from the reactor can be used for NPP auxiliary needs backup in case of a complete blackout. Thus, increased safety of NPPs will be ensured by installing an autonomous hydrogen energy complex. Moreover, this can reduce the risk of an accident when using both the newly developed and existing safety systems. This work presents an analysis of the current state of hydrogen production at NPPs, including the research results and rationale for efficiency of the developed energy complex. The calculations showed that development of NPPs in combination with autonomous hydrogen energy complexes will increase economic attractiveness of newly designed NPPs and will increase the competitiveness of existing nuclear plants on electricity market, will also provide their survival in the energy sector in the long term.

Model prediction of radioactivity levels in the environment and food around the world's first AP 1000 nuclear power unit.

Model prediction of radioactivity levels around nuclear facilities is a useful tool for assessing human health risks and environmental impacts. We aim to develop a model for forecasting radioactivity levels in the environment and food around the world's first AP 1000 nuclear power unit. In this work, we report a pilot study using time-series radioactivity monitoring data to establish Autoregressive Integrated Moving Average (ARIMA) models for predicting radioactivity levels. The models were screened by Bayesian Information Criterion (BIC), and the model accuracy was evaluated by mean absolute percentage error (MAPE). The optimal models, ARIMA (0, 0, 0) × (0, 1, 1)4, and ARIMA (4, 0, 1) were used to predict activity concentrations of 90Sr in food and cumulative ambient dose (CAD), respectively. From the first quarter (Q1) to the fourth quarter (Q4) of 2023, the predicted values of 90Sr in food and CAD were 0.067-0.77 Bq/kg, and 0.055-0.133 mSv, respectively. The model prediction results were in good agreement with the observation values, with MAPEs of 21.4 and 22.4%, respectively. From Q1 to Q4 of 2024, the predicted values of 90Sr in food and CAD were 0.067-0.77 Bq/kg and 0.067-0.129 mSv, respectively, which were comparable to values reported elsewhere. The ARIMA models developed in this study showed good short-term predictability, and can be used for dynamic analysis and prediction of radioactivity levels in environment and food around Sanmen Nuclear Power Plant.

The Impact and analysis of nuclear power plant’s steam extraction and heating modification on the turbine

A certain nuclear power unit uses the circulating water heated by the steam extraction from the high-pressure cylinder exhaust pipe of the steam turbine through modification, and the pressurized heating circulating water is supplied to the outside of the plant. As it is the first large-scale model of steam extraction heating in domestic nuclear power, the heating will have a certain impact on the steam turbine. Therefore, it is necessary to analyze the impact of large-scale steam extraction on the steam turbine and propose corresponding countermeasures to provide relevant references for subsequent nuclear power extraction and heating modification.

Effects of fluid dynamics parameters on flow-accelerated corrosion at elbow of carbon steel pipeline

Flow-accelerated corrosion (FAC) has always posed a significant threat to the safe operation of the secondary circuit in nuclear power units. In this study, we investigated typical carbon steel elbow pipe sections susceptible to FAC failure using fluid dynamics software to analyze the hydrodynamic characteristics at varying inlet velocities (2 m s−1, 4 m s−1, and 6 m s−1). The distribution of the FAC rate was monitored in real time using an array electrode. The results revealed that the outermost side of the elbow pipe section was the most susceptible location to FAC. By comparing different fluid dynamic parameters with the FAC rate, we identified radial velocity as an effective parameter for characterizing the FAC rate. Additionally, we established an empirical formula for predicting flow-accelerated corrosion in elbow pipe sections using the least squares method. The implications of this research are pertinent to the design and operation of pipelines in nuclear power plants.

The Application of Laser-Scanning 3D Model Reconstruction Technology for Visualizing a Decommissioning Model of the Heavy Water Research Reactor

Currently, over 100 nuclear power units globally have been in operation for more than 40 years. Hindered by the limitations of computer technology at the time, these nuclear facilities lack detailed electronic drawings. Activities such as equipment replacement and process circuit system modifications during operation result in discrepancies between paper drawings and actual conditions. Given the complexity and irreversibility of nuclear facility decommissioning activities, virtual simulation technology is often employed before the decommissioning process begins to assist in designing and validating decommissioning plans. Consequently, the creation of high-precision 3D models is crucial for subsequent decommissioning designs. Through innovatively utilizing laser-scanning 3D model reconstruction technology in the reconstruction of the model of China’s first heavy water research reactor undergoing decommissioning, this paper provides an overview of the process of laser-scanning 3D model reconstruction and its application in reconstructing the heavy water research reactor model. Using a 3D laser scanner, four decommissioning areas of the heavy water research reactor, including the reactor building, secondary water pump room, ventilation center, and low-level radioactive wastewater storage tank area, were subjected to 3D laser scanning. The acquired point cloud data from 572 scanning stations were processed using point cloud processing software for denoising, stitching, and triangulation. The triangulated model was then imported into modeling software for 3D reconstruction, ultimately establishing a digitalized model of the heavy water research reactor suitable for subsequent decommissioning simulation and design.

Status and future prospects of nuclear industry development in Bulgaria

This paper presents a study of nuclear energy development in Bulgaria and its contribution to the country's energy sector. Current status and future prospects of the nuclear power industry have been discussed.Currently, two 1000 MW nuclear power units – Units 5 and 6 with WWER-1000 reactors are in operation at Kozloduy NPP site. Nuclear energy is one of the main pillars of the electricity producing sector with 32, 6 % of current national energy mix.Energy strategy and vision for the development of the power sector in Bulgaria envisage nuclear energy to be supported institutionally as a promising resource for generation of emission-free electricity and to ensure a sustainable energy mix. Extending the operational lifetime of the available capacities, as well as, the construction of new nuclear capacities have been considered to be of strategic importance for preserving energy security, reducing greenhouse gas emissions and reducing dependence on imports of energy resources in a cost-effective and competitive manner.In addition, this investigation is focused on the efforts that have been invested toward current nuclear reactors lifetime extension, thermal power uprate and diversification of fresh nuclear fuel supplies, as well as, construction of new nuclear capacities to ensure the future development of the nuclear industry in Bulgaria.Some key drivers and challenges in nuclear power sector for future development were analysed.Large size nuclear reactors and SMRs in Bulgarian context were considered and the outcomes indicatesthat SMRs and large reactors generation III/III+ should not be opposed but SMRs could be considered as a complementarity technology to the existing large-scale reactors.Many of the currently developed SMRs conceptual designs meet the expectations for Generation IV reactors safety level but in-depth knowledge of the specific technology is required to assess and ensure safety in accordance with the Bulgarian regulatory framework which is oriented towards light water reactors.

Deep dynamic high-order graph convolutional network for wear fault diagnosis of hydrodynamic mechanical seal

The hydrodynamic mechanical seal (HDMS) in the reactor coolant pump of third-generation nuclear power units is vulnerable to failure due to prolonged operational periods and inevitable wear. However, traditional fault diagnosis methods are not robust to noise and can not leverage both the topological relationships among samples and local features. To resolve these challenges, in this paper, we propose a novel graph convolutional network (GCN) for wear fault diagnosis of HDMS called deep dynamic high-order graph convolutional network (DDHGCN). A dynamic graph learning module is designed to control the connectivity and sparsity of the iterated graph and thus eliminate errors and redundancies caused by noise. A high-order GCN module is proposed to effectively model the correlations between nodes, capturing contextual information and mutual influences among them. A residual convolutional module is applied to extract local features hidden in individual samples to further improve the classification performance. All three modules are jointly optimized for reliable wear fault diagnosis of HDMS. Experimental results demonstrate that our DDHGCN can achieve higher performance when compared with the state-of-the-arts.

Study of Load Adjustment Strategy for Nuclear Power Units Focusing on Rankine Cycle: Flexibility–Environment–Economy

The demand for the power grid system’s capacity to undergo peak-shaving is increasing as the proportion of renewable energy rises. In China, nuclear power units usually only provide a base load operation in the view of safety and economic considerations, but they do not provide load adjustment services, which undoubtedly increases the pressure of grid load adjustment. In this paper, a novel flexibility load adjustment strategy of the CHP nuclear unit is studied, which is achieved by introducing the thermal storage tank (TST) into the Rankine cycle without changing the output of the nuclear reactor. The AP1000 pressurized water reactor nuclear power unit for combined heat and power is taken as an example, and the thermodynamic model is established through the water vapor equation. Furthermore, the reference system is simulated for the goal of minimizing the imbalance between power supply and demand, and the flexibility–environment–economy benefits are evaluated. The results show that the heat storage/release of the TST may achieve power output flexible adjustment of the nuclear unit, and the power imbalance of the reference energy system is reduced from 1107.99 MWh to 457.24 MWh, a reduction of 58.73%. The introduction of a 600 MWh TST can enable the reference unit to contribute 335 MWh of peak electricity within the reference day. From the perspective of replacing the power generation output increment of coal-fired power units with equal amounts, it can achieve a reduction of 106.09 tons of coal consumption in the case day, which means that 277.73 tons of CO2 emissions can be reduced. The profit of the reference unit can be improved by CHY 70,125 via participating in load adjustment in the case day if following the time-of-use electricity price.

Intensification of Heat Transfer during Steam Condensation in Process Condenser of NPP Unit Cooling System

The paper investigates heat transfer during condensation of water vapor on vertical pipes of the process condenser of the cooling system of a nuclear power unit. The numerical study was performed at different mass flow rates of steam in the range of its change from 20 kg/s to 40 kg/s. Intensification of heat exchange is provided by the use of highly efficient heat exchange profiled tubes. The study used a set of heat exchange tubes (25 mm in diameter and 1.4 mm wall thickness) with the following values of the distance between the grooves of the profiled tube: 0.007075 m, 0.00875 m, 0.00925 m, 0.0105 m. The effect of the groove depth (from 0.0007 to 0.0009 m) on heat transfer during water vapor condensation on vertical pipes was also studied. The study was carried out for the range of changes in the Reynolds number for the condensate film from 5254.2 to 10508.5. The study obtained data indicating an increase in the heat transfer coefficient on the pipe with an intensifier compared to the heat transfer coefficient on a smooth pipe. This analysis did not take into account the change in tube wall temperature.

Research on the Multi-Angle Seismic Spectrum Response Law of Auxiliary Feedwater Electric Pumps

The auxiliary feedwater electric pump, which is vital in nuclear power units, demands exceptional seismic reliability. The impact of the seismic excitation direction on the seismic response of the complex auxiliary feedwater electric pump structure remains inadequately understood, thereby introducing uncertainties into seismic analysis and equipment installation procedures. To address this challenge, this study centers on a representative horizontal multi-stage centrifugal pump based on the finite element model and modal analysis. Utilizing the response spectrum method and the square root of the sum of the squares (SRSS) vibration pattern superposition principle, the research comprehensively explores the seismic response characteristics of individual pressure-bearing components and key rotor positions under seismic excitation from various angles. Findings reveal intricate variations in maximum stress and displacement responses for each pressure-bearing component under safe shutdown earthquake (SSE) level seismic excitation, corresponding to different input angles of seismic spectra. In assessing the seismic operability of the clearance between the impeller and the stator, the direction near 45° exhibits the maximum displacement response, highlighting the need for focused attention in testing and verification checks. The methodologies and conclusions presented in this paper offer valuable insights for designing, optimizing, and installing horizontal multi-stage centrifugal pumps, including auxiliary feedwater electric pumps, providing valuable guidance for future applications in the field.

Method for Creating Mathematical Models of Heating Nuclear Power Units of Power Plants for Optimization Studies of Autonomous Electric Power Systems

The authors present a two-stage method for creating mathematical models of cogeneration nuclear power units for optimization studies of autonomous electric power systems. The first stage of the methodology includes the development of a detailed model of a nuclear power unit, which provides a satisfactory accuracy of the description of physical and technological processes, as well as optimization calculations of operating modes. At the second stage, the energy dependencies are built based on the results of optimization calculations of the first stage. Energy dependencies define the boundaries of the region of feasible solutions in the form of polynomials. A simplified mathematical model of a nuclear power unit for optimization studies of autonomous EPS is created using polynomials. To solve the problem of finding a polynomial, a two-step approach is proposed. At the first step, the polynomial coefficients are selected such that the minimum of the maximum value of the modulus of the difference between the function determined using the polynomial and the function determined using the detailed model of the power unit is reached. At the second step, the deviation modules are limited to the value found at the first step, and the sum of the deviation modules is minimized at all points. The two-stage technique developed by the authors is demonstrated on the example of a cogeneration nuclear power unit, which is supposed to be operated in the climatic conditions of the Far North.

Suitable Analysis of Micro-Increased Capacity Model on Cold-End System of Nuclear Power Plant

The cold-end system of a nuclear power plant is a key complex node connecting the power generation system with the variable environmental conditions, and its operation, economy, and stability have become the main obstacles to further improving the performance of the first and second circuits. The current research on the interactions between the cold-end system and the thermal cycle of nuclear power mainly adopts the micropower model, while the existing condenser model does not take into account the influence of the turbine exhaust resistance and exhaust flow and other factors on the condenser vacuum change caused by the change in the circulating water flow rate and temperature in determining the optimal vacuum. This ignores the interactions between the equipment and the interconnections between the parameters, which results in the reduction of the model’s accuracy. This paper takes a nuclear power unit as an example, adopts the “constant flow calculation” method to calculate the heat balance of the two-loop thermal system of the nuclear power plant, and constructs an integrated simulation model of the reaction environment variables, the cold-end system, and the thermal cycle. Taking the circulating water temperature and flow rate as variables, the errors of the separate condenser model and the coupled model in circulating water parameter changes were obtained under the condition of satisfying the thermal system operation, and the circulating water temperature and flow rate change ranges applied by the separate condenser model were analyzed in order to reduce the amount of calculations when the unit power error was 1%. The results show that the circulating water temperature is 4 °C, the applicable range of the circulating water flow rate is 42 m3/s to the rated flow rate, the applicable range of the circulating water temperature is 20 °C, the applicable range of the circulating water flow rate is 32.12 m3/s to the rated flow rate, the applicable range of the circulating water temperature is 26 °C, the applicable range of the circulating water flow rate is 38.63 m3/s to the rated flow rate, the applicable range of the circulating water temperature is 30 °C, and the applicable range of the circulating water flow rate is 45 m3/s to the rated flow rate. At a circulating water temperature of 26 °C, the applicable range of the circulating water flow is between 38.63 m3/s and the rated flow; at a circulating water temperature of 30 °C, the applicable range of the circulating water flow is between 45.64 m3/s and the rated flow.

Modeling and operation strategy of nuclear power plant with electric heat storage in the ancillary service market

In the power ancillary service market, power sources can participate in peak shaving and obtain profits. However, nuclear power plants (NPPs) cannot contribute sufficient peak-shaving capability to the power system and obtain peak-shaving profits due to their limited flexibility as the base load. To tackle this challenge, an electrothermal coupled model for NPPs is proposed in this paper, which combines the characteristics and constraints of nuclear power units and electric heat storage (EHS). The coupling model can enhance the flexible adjustment capability of NPPs by increasing their self-consumption. Meanwhile, the operation strategy of NPPs considering the electrothermal coupling model is developed to enhance the NPP operation profits. Moreover, a cost calculation method for NPPs is developed, which can calculate the peak-shaving compensation and cost-sharing of NPPs. The effectiveness of the proposed method is verified by numerous simulations using the model of the Hongyanhe Nuclear Power Plant in Liaoning, China. Compared with the three traditional methods, the profit of the EHS is increased by 3%, the peak-shaving compensation is enhanced by 4.4%. Meanwhile, the cost-sharing is reduced by 40%, and the total profits of the proposed method are improved by 18.9%.

Perspectives of nuclear energy development in Ukraine on the global trends basis

The article examines global trends in the development of the nuclear power industry. It includes the following: extending the operating life of nuclear power units; development of atomic energy in the context of the Paris Agreement; development of nuclear-hydrogen energy; synergistic interaction of renewable energy sources and nuclear power plants; introduction of new reactor technologies. The attitude of different countries of the world to atomic energy is described. It is determined that China and India are the leaders in developing nuclear power. It was determined that small modular reactors are considered transformative reactors that will contribute to the further development of atomic energy in the world. The advantages of small modular reactors in comparison with reactors of large capacity are described, and recommendations for selecting small modular reactors for Ukraine are formulated. Installation of small modular reactors at the operational sites of the NPPs of Ukraine can reduce the financial costs of their construction. Therefore, it will contribute to the sustainable development of the nuclear energy industry of Ukraine.

Numerical Simulation of the Loss of Offsite Power to the VVER-1200 Nuclear Power Unit Auxiliaries

Numerical Simulation of the Loss of Offsite Power to the VVER-1200 Nuclear Power Unit Auxiliaries

Analytical Technology of the “Economic Cross” — a Toolkit for Modeling the Effectiveness of the Nuclear Fuel Cycle Closure

Changes in the economic development priorities during transition to a post-industrial society inevitably forces us to reconsider approaches to the role of innovation in the modern economy, including in nuclear energy. The life cycles of nuclear energy resources and nuclear power units are intersected, and at this intersection energy is generated, revenue occurs, etc. The rest of these life cycles are mainly costs. Method of analyzing this life cycles intersection was named the “economic cross”. The purpose of the study presented in the article is to specify conditions, under which the final revenue, with regard to discounting, will exceed the total costs, and nuclear power technology will be profitable.

Mechanical and radiation shielding characterization of W-based alloys for advanced nuclear unit

The paper aims to investigate the mechanical and radiation shielding properties of two tungsten-based alloys manufactured by powder technology; their features when compared with two standard stainless steel grades for advanced nuclear applications. Multiple measurements were performed to characterize the alloys' structural and mechanical properties. XRD analysis and average surface roughness measurements showed the crystalline and morphological structure of the alloys. Surface microhardness added to the abrasive wear analysis showed the final wear resistance of the selected alloys. In addition, the shielding features against gamma and fast neutrons radiation were calculated. The superior characteristics of W-based alloys manufactured by powder technology make them suitable to be used in advanced nuclear power units.

Rotating Rectifier Fault Diagnosis of Nuclear Multiphase Brushless Excitation System Based on DTW Metric and kNN Classifier

Multi-phase annular brushless excitation system is widely used in large-capacity nuclear power units. Accurate fault diagnosis of the rotating rectifier is of great significance to improve the reliability of the excitation system. However, the traditional diagnosis method based on harmonic analysis of stator field current has some deficiencies. In this paper, a novel diagnosis method using field current waveforms and artificial intelligence is presented. Firstly, the various shape features in field current waveforms of the different rotating rectifier faults are analyzed. Then, the field current waveforms are used as the input of a hybrid algorithm based on the dynamic time warping (DTW) metric and the <italic xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">k</i> -Nearest Neighbors ( <italic xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">k</i> NN) classifier (DTW- <italic xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">k</i> NN). That is, a DTW metric is used to calculate the distance among the shape features in field current waveforms and <italic xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">k</i> NN classifier is used to diagnose the specific rotating rectifier fault. Finally, experiments on an 11-phase prototype prove the effectiveness of the hybrid method DTW- <italic xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">k</i> NN. It is worth mentioning that an improved training set including all trends of field current waveforms should be selected to avoid the asymmetry between each pair of field poles. The learning method provides a new idea for fault diagnosis of the rotating rectifier.