Operation Of Nuclear Power Plants Research Articles

Performance influence of polyacrylic acid dispersant on nuclear grade anion exchange resin

Secondary circuit water chemistry control is vital for the safe operation of nuclear power plants. Polyacrylic acid (PAA) is used in the secondary circuit water system to reduce the fouling of the steam generator (SG) because of its good dispersion effect. However, PAA as a macromolecular organic matter possibly causes organic fouling for anion exchange resin of the water purification system and affects its performance. Some research reported that the performance of resin changes after PAA application. But it is not clear whether this change is related to PAA. In this research, the anion resins were performed for five cycles, and each cycle included batch PAA adsorption, dynamic PAA desorption and resin rinse. Results showed that NaOH solution could realize the desorption of PAA on anion resin, but desorption solution and rinse water consumption slightly increased with the cycle number. Meanwhile, residual PAA linearly increased with the cycle number, but the rate was relatively slow. Furthermore, the performance influence of PAA on anion resin was investigated. The exchange capacity and kinetic performance for anion resin slightly decreased after PAA cycle. The effluent conductivity of anion resin bed increased after PAA cycle, but it could not cause the adverse influence of sodium leakage. The contrast experiment of H2SO4 cycle showed that resin aging was the primary reason for the change of resin performance, rather than residual PAA. Within the fewer cycles, PAA and anion resin had better compatibility.

Application of EDG AOT Extension Based on the Risk-Informed Method in NPPs

At present, the allowed outage time (AOT) of an M310 unit emergency diesel generator (EDG) was 3 days, which can be extended to 14 days through replacement of additional diesel units; although it provides a certain online maintenance time, it cannot meet the needs of ten-years overhauls. In order to avoid stopping the reactor for maintenance of NPP due to insufficient of EDG AOT, based on risk-informed method analysis feasibility of extending AOT for EDG to 28 days, we quantitatively calculate the impact of extension of AOT on risk level of nuclear power plants (NPPs). Analysis shows that extension of EDG AOT to 28 days has less impact on NPPs, and safety of NPPs can be further ensured through temporary risk control measures, so the extension of AOT to 28 days is acceptable. By using risk-informed technology to extend AOT for EDG, unnecessary shutdown and maintenance is avoided and the economy of NPPs and flexibility of maintenance work arrangement is greatly improved while ensuring safety, which is of great significance to operation and maintenance of NPPs.

Influence of climate data quality on predictive accuracy of production and consumption of energy resources by a city

The article provides a list of hazardous climatic phenomena and examples of some of their effects on the constituent structural elements of the power plant. In the framework of this article, the impact of the duration of hot weather on the operation of nuclear power plants is considered. An analysis was made and the stations most susceptible to global climate change in terms of increasing summer temperatures were identified. For the two most representative stations, modeling based on artificial neural networks was performed and conclusions were drawn about an increase in the number and duration of hot days per year. The best forecast result for St. Petersburg NPP is 54.59%, for Kursk NPP - 22.18%. The structure and method of formation of the predictive function are analyzed. Methods are proposed to improve the accuracy of the forecast by varying the discretization value of the initial data, as well as by changing the activation function of the artificial neural network.

МЕТОДИКА ОБОСНОВАНИЯ КРИТЕРИЕВ И ПОКАЗАТЕЛЕЙ ОЦЕНКИ ЭКОНОМИЧЕСКОЙ УСТОЙЧИВОСТИ АТОМНОГО ЭНЕРГОПРОМЫШЛЕННОГО КОМПЛЕКСА

The article describes the methodology for substantiating the criteria for assessing the economic sustainability of the nuclear power industry in a competitive energy market. A classification model of groups, types of criteria and characteristics of sustainability indicators has been formed. Substantiation of evaluation criteria and parameters is an important task for ensuring the sustainable functioning of the industry and ensuring the safe operation of nuclear power plants.

Graph classification based fault detection in nuclear power plants with graph formulation

In the routine production and operation of nuclear power plants (NPPs), safety management is extremely important, and consequently, a fault detection system yielding high accuracy could provide effective decision support and prevent the accidents to the maximum extent. Most of existing studies are based on the analysis of time-series data, while researches oriented to the implementation of graph data mining are barely seen. In this paper, we propose an improved fault detection method to address this task from the perspective of graph representation learning. By calculating the similarity between indicators, the multivariate temporal sequences of the running characteristics of NPPs have been formulated as a group of graphs with corresponding labels, on which the graph convolutional network (GCN) has been introduced to detect fault samples in the way of graph classification. Compared to conventional supervised learning classifiers, our proposed methodology performs more superior identification of major faults, and the sensitivity regarding significant parameter or settings in the model design has been analyzed to investigate their optimal arrangements. The findings in this article not only present the promising prospects of introducing graph learning techniques into safety management, also indicate that various kinds of digital signals could be reformulated into graph dataset based on which the original task could be executed more effectively in the way of graph representation learning.

More Efficient Heat Recovery from the Condensate of Reheaters at New and Operating Nuclear Power Plants (NPPs)

More Efficient Heat Recovery from the Condensate of Reheaters at New and Operating Nuclear Power Plants (NPPs)

Flow methods as a general approach to automation of the chemical analysis of aqueous media (review)

The article is devoted to the consideration of general approaches to automating the analysis of aqueous media, which are the main objects of chemical analysis including natural and waste waters, as well as technological media in various industries (microelectronics, biochemical industries, nuclear and thermal power engineering), which are the main consumers of high-purity water used for pre-start flushing of inline equipment to the required cleanliness of surfaces, for preparation of coolants that ensure the operation of nuclear and thermal power plants. In all these cases, the content of impurities in used or spent and intended for discharge aqueous media is strictly regulated. Changes in the content of impurities in controlled media often occur spontaneously due to equipment failure on the controlled objects. Prompt response to such changes and their timely elimination can be performed more efficiently if chemical and technological control of the normalized parameters of aqueous media of any named origin and destination is carried out in a continuous automatic, on-line, mode, which guarantees the exclusion of gaps or delays in obtaining information about equipment failure on a controlled facility in case of technological control or about unauthorized discharges of pollutants in case of environmental control. Possible methodological approaches to the development of systems for continuous chemical monitoring of quality or levels of contamination of aqueous media are considered on the basis of the above prerequisites, and applicability of various flow methods for these purposes is discussed in the order of their appearance in the arsenal of analytical chemists.

Порівняльний метод кваліфікації систем безпеки ядерних енергоустановок з ВВЕР‑1000 та АР1000

A deterministic method of analyzing conditions for ensuring the safe operation of VVER-1000 and AP1000 nuclear power plants has been developed. The method is based on the comparison of the provision of the required safety functions by critical (minimum permissible) configurations of safety systems for accident management. During the analysis, the composition and information on the technical characteristics of the safety systems of the VVER-1000/320 and AP1000 reactor installations, which are designed to ensure the safety functions of feeding and cooling the reactor/PG, were considered. On the basis of the developed method, the relative criteria and safety conditions (for example — the criterion of safety — the maximum temperature of the heating element shell; the condition of safety is not to exceed the maximum design limit of damage to the twelfth (the temperature of the twelfth shell is 1,200 °C) for nuclear power plants with VVER-1000 and АР1000 reactors in terms of safety functions for emergency fueling and cooling of the reactor core and steam generators are determined. Preliminary estimates of the relative criteria and conditions for ensuring nuclear safety during accident management of natural circulation circuits of passive safety systems AP1000 obtained taking into account data on their composition and technical characteristics, the table showed that the required height of natural circulation circuits of passive safety systems should be several hundred meters. Based on the established relative criteria and safety conditions, the need for additional modernization of the АР1000 safety systems to successfully ensure the safety functions of emergency coolant/feedwater supply and core and steam generator cooling is preliminarily determined.

INFLUENCE OF PRESSURE IN THE TURBINE CONDENSER ON HEAT SUPPLY EFFICIENCY OF NPP WITH HEAT PUMP

The use of a heat pump for heat supply makes it possible to practically stop thermal pollution of the environment during the operation of thermal and nuclear power plants in winter. If a steam turbine condenser is used as a low-potential energy source for heat pump, the amount of released thermal energy will be equal to the sum of the thermal power of the NPP condenser and the power of the heat pump compressors. From the point of view of environmental safety, heat supply by combining power plant with a heat pump is an urgent task. But it is known that the due to the lack of steam extraction for water heating, the additional electrical power of the cogeneration heat and power plant will be less than the capacity of the heat pump compressors. Thus, in terms of thermodynamic efficiency, the use of a heat pump loses to a traditional cogeneration plant. The purpose of the work is to determine the influence of the final pressure in the turbine condenser on the thermodynamic efficiency of a nuclear power plant with a heat pump. A mathematical model of the thermal scheme of the K-1000-5.8/1500 NPP turbo-plant during summer and winter operation with heating plant has been developed. With the heating plant capacity of 230 MW, the electric capacity of NPP unit decreases by 43.5 MW. A mathematical model of a heat pump has been developed, for which a steam turbinecondenser is used as a low-potential energy source. To ensure the release of 230 MW of heat, the power of the heat pump compressor must be 48.4 MW. Thus, if the heating plant is replaced with a heat pump of the same capacity, the electric power will decrease by 4.8 MW. Calculations were made regarding the influence of the final pressure in the condenser on the exergetic efficiency of the NPP with heat pump, which uses the entire capacity of the turbine condenser. The analysis of the obtained results showed that the exergetic efficiency due to the increase in electric power released in winter increases with the increase of the final pressure in the condenser. This is explained by an increase in the heat pump coefficient of performance.

Waste-specific volume for retrospective, predictive assessments, and ranking of practices during normal operation of Russian nuclear power plants

Possibilities of applying the generated radioactive waste-specific volume per unit of produced electricity are shown. This waste specific volume is used for retrospective assessment and forecasting of radio active waste volumes generated at Russian nuclear power plants. According to the available data period covering 2008-2021, the mean and median values of the annual waste-specific volume for each nuclear power plant were obtained. The medians for solid radioactive wastes divided into the categories of very low-level wastes, low-level wastes, intermediate level wastes and high-level wastes are equal to 3.6?10-2, 3.2?10-2, 3.2?10-3, 3.0?10-4 m3(GWh)-1, respectively. For liquid radioactive wastes of the low-level waste and intermediate level waste categories - 1.3?10-3 m3(GWh)-1, 2.4?10-2 m3(GWh)-1, respectively. The highest mean and median values of waste-specific volume for all radioactive waste categories are typical for nuclear power plants with LWGR (RBMK) reactor installations. The forecast based on the plans to increase electricity production by Russian nuclear power plants indicates a likely increase in the volume of radioactive waste generation by 0.7-7.4 % (depending on the waste category) in the period from 2022 to 2027. The waste-specific volume use makes it possible to rank the existing practices of nuclear power plant operation by the volume of radioactive waste generation to justify the criteria for compliance with the International Project on Innovative Nuclear Reactors sustainability methodology.

A point cloud semantic segmentation method for nuclear power reactors based on RandLA-Net Model

The detection of foreign objects in nuclear power plant reactor is a key task in the operation and maintenance of nuclear power plants. Loose and falling foreign objects such as bolts can lead to fuel component damage and unplanned shutdown, posing serious hazards. Therefore, we propose a point cloud semantic segmentation method for foreign objects in nuclear power plant reactor based on the RandLA-Net model. Considering the correlation between point cloud collection error and curvature, the data augmentation method is improved to reduce the risk of model overfitting. By treating the boundary points of different classes as the hard examples, the hard example mining is designed to improve model generalization performance. Adding an improved test time augmentation method during model inference, the more reliable segmentation results are performed by multiple prediction on points. The experimental results indicate that the proposed method can achieve high-accuracy reactor point cloud semantic segmentation with mIoU of 0.992 and mAcc of 0.997.

Результати діяльності Інституту проблем безпеки атомних електростанцій НАН України у 2022 р.

The most important results of the scientific and scientific-organizational activities of the Institute for Safety Problems of Nuclear Power Plants of the National Academy of Sciences of Ukraine (ISP NPP, NAS of Ukraine) in 2022 are presented. It is shown that even under the extremely difficult conditions in which the entire country is due to the invasion of Russian troops into the territory of Ukraine, the employees of the Institute made great efforts to implement planned tasks to ensure nuclear and radiation safety of the destroyed fourth power unit of the Chornobyl nuclear power plant, handling of nuclear hazardous materials, as well as increasing the reliability and safety of nuclear power plants operation. The results of these studies were implemented in the fields of production, electric power industry and during the educational process, as well as highlighted in a number of publications and scientific events. Thus, despite the objective difficulties and guided by the development strategy, the ISP NPP plans to preserve the scientific and technical potential under the conditions of power outages, lack of heating and water supply in some places, disruption of the functioning of public transport, telecommunications and data transmission facilities, etc., as well as to become the leading organization of Ukraine for scientific and technical support of activities for the peaceful use of nuclear energy

Codes and methods improvements for safety assessment and LTO: varied approaches

Nuclear safety has always been at the heart of the concerns of nuclear power plant operators and developers, as well as of various nuclear research organizations and regulatory authorities. Over the last decades, all these nuclear actors have developed and integrated a large number of calculation codes and other tools into their safety work. From the system approach to the local understanding of a phenomenon on a given component, from neutronics to operation optimization for long-term operation, these methods and codes have been constantly evolving since their appearance, in order to be able to integrate new plant designs and components, to improve the results of modeling physical phenomena or quantify and thus reduce the uncertainties on these results. Currently, several H2020 Euratom projects are working on the improvement of these codes and methods. This article will focus on three of these projects: CAMIVVER (Codes And Methods Improvements for VVER comprehensive safety assessment), APAL (Advanced PTS Analysis for LTO), and sCO2-4-NPP (innovative SCO2-based heat removal technology for an increased level of safety of Nuclear Power Plants) in order to illustrate our thinking on the improvement of calculation frameworks. First, we will present the work and the approach adopted with regard to the different calculation codes and methods used in each of these three projects. We will then conclude with an overall analysis of these three approaches, highlighting the difficulties and successes of these three projects, and identifying areas of work for the general improvement of the calculation codes.

Role of HMS Research Division for resumption of nuclear power plant operations (2)

Role of HMS Research Division for resumption of nuclear power plant operations (2)

Role of HMS Research Division for resumption of nuclear power plant operations (1)

Role of HMS Research Division for resumption of nuclear power plant operations (1)

가상경계법과 중첩격자법을 활용한 스윙 체크 밸브의 동수력학적 특성 비교

Nuclear Power Plant operators regularly check the performance of pumps and valves to verify the safety functions, and to monitor the degree of vulnerability over time through In-Service Test(IST). The swing check valve is one of the representative IST-related valve type. This valve functions to flow the fluid in only one direction, and automatically closes when the backflow occurs. In this study, the swing check valve performance test is conducted, and the result is compared with the computational fluid dynamics analysis using the immersed boundary method(ANSYS CFX R19) and overset mesh method(ANSYS Fluent R19). Overall, the inlet pressure depending on the disc angle predicted by the Fluent(overset mesh method) showed less error than the measured data compared to the CFX(immersed boundary method).The difference in the flow pattern passing through the gap between the valve entrance and top of the valve disc predicted by both methods was estimated to be the cause of the error in the simulation result compared to the measurement.

Construction of the Software and Technical Complex of Control Basic Parameters of Reactor Installation

The main problem associated with the operation of nuclear power plants is the problem of correct and timely diagnosis of failure or violation. The mistakes of technologists can lead to severe damage to nuclear power plants, or simply to a reduction in the utilization rate of installed capacity. The personnel make the main mistakes in the conditions of time shortage and being in a stressful situation during the development of an accident, when assessing changes in emergency parameters is not always possible, which leads to incorrect diagnosis of initial events (IE). Any IE leads to the deviation of the monitored parameters from normal values (values that are inherent to normal operation). The operator-technologist needs as soon as possible to determine the parameter (or parameters) for which the change occurs and, by performing a certain sequence of actions, bring their values to the norm. For information support of the operator-technologist is the information computing system (ICS). In the ICS, it is possible to control all parameters affecting the safe operation of the equipment, for each of which there are certain values, the deviation from which may lead to an accident. The paper considers the issues of the need to separate the measured parameters of NPP into separate groups that uniquely characterize the state of the critical security features. In particular, the necessity of consideration as the most important mass parameter of the primary coolant is shown. The issues of level measurement problems in the pressure compensator are discussed as the most important from the point of view of determining the mass of the primary coolant. Methods are proposed for determining the operability of sensors, the method of calculating a reliable level in the volume compensator. Estimates of the computational efficiency of the proposed methods are given.

Simulation of Dispersion Modeling of 137Cs for the Possible Leakage of Malaysia’s Nuclear Power Plant Operation

The world’s energy needs increase in line with population growth. One alternative to overcome this problem is the construction of a nuclear power plant, a source of energy that is cheap, clean, and safe. Malaysia has a plan to construct nuclear power plants by 2025, which is located close to Indonesian waters. This study aimed to determine the distribution model of 137 Cs radionuclide released by the nuclear power plant in the event of a leak in east coast of Peninsular Malaysia and its impact on the presence of 137 Cs in Indonesian waters. A quantitative method was used in this study with a scenario 2D modeling using hydrodynamics module and transport module in MIKE software. The results showed that the highest concentration of 137 Cs would be found in the area around the nuclear power plant with a value of 10 11 PBq/m 3 , then it would decrease through diffusion and advection processes. On the 15 th day, the spread of 137 Cs would reach Indonesian waters with a distance of up to 76 km and would expand on the 30 th day with a distance of up to 130 km from the released source. The movement of 137 Cs follows the dominant current pattern due to its nature. The presence of 137 Cs in Indonesian waters after the 15 th and 30 th days would reach a value of 1 Bq/m 3 due to advection and diffusion processes.

Design and evaluation of ecological interface for Feedwater Deaerating Tank and Gas Stripper System based on cognitive work analysis

Abstract The investigation of nuclear power plants (NPPs) accidents in the past shows that most of the accidents mainly occur in unexpected events. In this study, in order to verify whether Ecological Interface Design (EID) improves the situational awareness of operators in NPPs, this paper first analyzes the system by using the first three stages of cognitive work analysis, and then applies EID to the operating interface of NPPs to develop an ecological interface. In order to make the test results more complete, an improved interface has also been developed. A process expert and six operators were invited to participate in our experiments to measure situational awareness. The results show that the situational awareness of ecological interfaces in unexpected events is significantly higher than that of traditional and advanced interfaces. The significance of this study is that EID, as a practical technology, can be widely used in operator control rooms to improve the ability of operators to solve unexpected events.

Study on radiation degradation of nuclear power cable materials

During the operation of nuclear power plants, it is necessary to monitor and evaluate the aging of cable materials. In this paper, we evaluate the irradiated use status of two types of cable insulation materials used in nuclear power plants-cross-linked polyethylene (XLPE) and ethylene propylene diene methylene (EPDM). Firstly, two kinds of cable materials are subjected to accelerated radiation aging with different cumulative doses, and the macro-mechanical properties and microstructure changes of the cable materials after different doses are studied. Analyze and test its hardness value, tensile strength, elongation retention at the break, scanning electron microscope, and oxidation induction temperature. The results show that after γ-ray irradiation, the cable hardness and the retention rate of elongation at break show a downward trend. Under the highest cumulative dose of 500kGy, the retention rate of elongation at break is greater than 50% of the failure index. The tensile strength decreases slightly, the microscopic morphology of the tensile section has holes, and the oxidation induction temperature decreases.