Economy Of Nuclear Power Plant Research Articles

In-reactor behaviour and economic assessment of enriched gadolinia burnable absorbers

The increasing demand for energy and rising energy prices have increased the importance of the efficiency and economic viability of operating nuclear power plants. One approach to improving the reactivity control in such cases is the use of burnable absorbers such as gadolinium oxide (Gd2O3). While Gd2O3 has proven effective in regulating reactivity in light water reactors, it has certain drawbacks, including the displacement of fissile isotopes from the fuel composition and residual reactivity penalties caused by isotopes other than the main neutron absorbers. In this study, the potential impacts of 157Gd-enriched Gd2O3 on the operation, safety, and economy of nuclear power plants were assessed. Fuel cycle analyses were conducted using the advanced nuclear design code system Studsvik CASMO4/SIMULATE3. Several aspects were analysed, including peaking factors, reactivity feedback parameters, power distribution, and shutdown margin. The results of the analysis revealed that by removing the isotopes responsible for residual reactivity from Gd2O3, it was possible to achieve the same fuel cycle length with lower uranium enrichment levels, enabling a higher concentration of fissile isotopes and yielding economic benefits without compromising safety.

Reliability Evaluation of Complex Nuclear Energy Redundancy Systems with Atypical Time-Distribution Events

In complex nuclear energy redundancy systems, there are many failure events that do not follow specific time distribution. For these atypical time-distribution events, traditional dynamic fault tree (DFT) methods cannot be applied directly, which has posed great challenges to reliability modeling and evaluating. In this contribution, we summarize atypical time-distribution events in nuclear energy redundancy systems and propose new modeling and evaluating methods based on DFT. To demonstrate the reasonability of the proposed methods, two case studies about make-up water pumps and emergency diesel generators are analyzed in comparison with traditional DFT. The results indicate that the proposed methods can effectively model and analyze the reliability of redundant systems with atypical time-distribution events. The proposed methods can provide useful information for optimization design of nuclear energy redundancy systems and has potential to improve the economy of nuclear power plants by relaxing overestimated unreliability.

Research and Application Practice on Technical Supervision of Nuclear Island Mechanical Equipment in NPP

The safe and reliable operation of nuclear island mechanical equipment in NPP is directly related to the stable power generation of units. Strengthening the supervision and management of nuclear island mechanical equipment is an important measure to improve its reliability. This paper presents a nuclear power plant nuclear equipment purpose and significance of electric power and technical supervision, technical supervision scope, equipment identification methods, technical supervision content build process, and the reactor coolant system main pump as an example, compiled the technical supervision guide, some CPR1000 nuclear power plant in the pilot application. The existing problems in the management of the main pump group were found, and suggestions for improvement, reliability improvement and economy improvement were put forward. The practice has proved that the technical supervision of nuclear island mechanical equipment in nuclear power plants is helpful to improve the power generation capacity and economy of nuclear power plants.

Study on ultrasonic attenuation characteristics in air-droplet two-phase flow

The wetness of exhaust steam in nuclear power plant steam turbine is an important index to evaluate the safety and economy of nuclear power plant and the ultrasonic attenuation method has the advantages of quick response and non-contact in the steam wetness measurement of nuclear power plant steam turbine. The saturated steam-droplet medium and the air-droplet medium have many similar thermal fluid characteristics, in order to simplify the experiment, the air-droplet two-phase flow is used to simulate the wet steam discharged from steam turbine of nuclear power plant, and the ultrasonic attenuation coefficient in the air-droplet two-phase flow is measured. The correlation between wetness and ultrasonic attenuation coefficient was studied by calibrating the wetness of the air-droplet two-phase flow and measuring the ultrasonic attenuation coefficient under different wetness conditions. Meanwhile, the static characteristics of ultrasonic transducer and the influence of flow velocity on attenuation coefficient are studied experimentally. According to the results of experiment and numerical calculation, the ultrasonic attenuation coefficient is very sensitive to the change of wetness. The ultrasonic attenuation coefficient and wetness has a good linear relationship in the wetness range (0–16%), the ultrasonic attenuation method is appropriate for the wetness measurement of turbine steam exhaust and has high precision when the droplet size is stable.

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.

NSGA-II algorithm-based LQG controller design for nuclear reactor power control

The design of reactor power control system is closely related to the safety and economy of nuclear power plants. To improve its power control performance, this study takes a pressurized water reactor as the research object. The nonlinear dynamic mathematical model and its state space model were first derived; Then the Linear Quadratic Gaussian (LQG) optimal control method was adopted to design the reactor power controller; Furthermore, the multi-objective optimization of the linear quadratic regulator (LQR) weighting coefficient is carried out based on the elitist nondominated sorting genetic algorithm version II (NSGA-II). The results show that the LQG controller optimized based on NSGA-II method not only has fast response, high control accuracy and good stability, but also has good robustness and can obtain superior reactor power control performance.

Modal and structural analysis on a main feed water regulating valve under different loading conditions

As a key device influencing the safety and the economy of nuclear power plants simultaneously, the main feed water regulating valve (MFWRV) controls the flow rate flowing to the steam generator. In this work, modal and structural analysis is conducted under different loading conditions to investigate vibration characteristics and mechanical characteristics of MFWRV. The results show that the combination of loading conditions has little effect on the natural frequency and the mode shape of MFWRV. However, it influences the results of structural analysis significantly. Compared with the fixed internal pressure, the variable internal pressure promoted the maximum stress intensity when the valve opening (L/Lmax) is smaller than 80%. The maximum deformation when the seismic load is considered is almost three times as large as it when the seismic is not considered. The seismic load always has a positive effect on the maximum gap (gmax) between the valve core and the sleeve. However, the variable internal pressure only has a positive effect on gmax relative to the fixed internal pressure when L/Lmax = 40 ~ 90%. Besides, the variable internal pressure has an evident negative effect on the maximum sliding distance (sdmax) between the valve core and the sleeve relative to the fixed internal pressure under the low and the middle valve opening. This work provides a reference for the structural design of the main feed water valve and other similar valves.

Best estimate plus uncertainty analysis of a small‐break LOCA on an advanced Generation‐III pressurized water reactor

To improve the economy of nuclear power plants, a passively pressurized water reactor with additional 40% power output was developed based on AP1000 technologies. In this study, the RELAP5 code was used to model the reactor system components involved in the accident transient, and a postulated 2-in. small break loss of coolant accident (SBLOCA) was simulated and analyzed. The evaluation of the code applicability was conducted for the development models and special process models. With the base input deck, a best estimate plus uncertainty (BEPU) analysis was performed. The significant contributing parameters to reactor safety criteria were identified according to the phenomenon identification and ranking table. The DAKOTA software was used for simple random sampling, and a set of 93 calculations were performed based on the Wilks' formula. Important input uncertainty contributors for the predicted minimum core collapsed level were identified through the spearman rank correlation coefficient and partial rank correlation coefficient.

Extenics matter-element analysis on dilemma problem in HMI design of nuclear power plant

The Human machine interface (HMI) plays a critical role for operator monitoring and controlling the system. The background colors of HMI, as one of human factors, directly influence on the performance of operator, including sensitive and fatigue degree, and furthermore affects the safety and economy of nuclear power plant (NPP). The selection between color differences (affects sensitive and number of matched colors) and performance overtime (affects fatigue degree) are dilemma problem. This paper proposes an extenics matter-element analysis to settle the dilemma problem in HMI design of NPP. By combining both matter-element analysis and extension in HMI, this method changes opposability problem into consistent problem, and discrete problem into continuum problem. A basic background color selection model and weighting method are established too. Taking HMI design as an example, this paper expounds a basic method of solving dilemmas by extension matter element theory. The model, which reflects the distance between selected element and the optimal interval that can be used and further developed for analyzing or solving similar dilemma problem, it is effectively find the best answer for human performance in industry design, and the design of NPP can be further optimized based on this method.

核电站凝水系统的状态预测方法研究

运行支持系统可以辅助操纵员进行决策分析，对于提高核电站的安全性和可靠性具有重要意义。而发生故障后的状态预测方法是目前运行支持技术中的热点问题，本文提出基于机理仿真的凝水系统状态预测方法。在进行故障特性分析的基础上，将从仪控系统中采集到的实际测量数据引入机理仿真过程中，在核电站正常运行时仿真模型能与之同步运行；当电站发生故障后，能快速完成异常的识别并切换至超实时运行进行运行状态预测。通过采集600 MW全范围仿真机中的正常和故障数据进行实验测试表明，本方法可以实现有效的状态预测，可以辅助操纵员评估故障的发展变化趋势和严重程度。 Operation support system can improve the safety, reliability and economy of nuclear power plant, condition forecasting technology is one of the hot research problems in operation support system. The major work of this thesis is researching the technology of condition forecasting of condensate water system. By a detailed analysis of the composition and the failure characteristics of the condensate water system, the paper uses the simulation model to predict the condition of the equipment. When the equipment is in normal, the system can simulate the current state in realtime; when the failure occurs, the condition forecasting module based on simulation model run in ultrareal time to predict the future states and anomalies of the equipment. The method has been proven by simulation validation. Results show that the condition forecasting system meets the requirements of predicting the state of nuclear power plants, and provides sufficient information for the operators.

AP1000屏蔽式主泵维修可行性分析

核电厂反应堆冷却剂泵是核安全一级设备，其运行的可靠性直接关系到电厂核安全、辐射安全及经济性，对防止核电厂事故的发生和发展极为重要，并且发生故障主泵的解体检修可能是核电厂大修期间的重要工作之一。AP1000反应堆主泵采用单级离心飞轮带屏蔽电机的屏蔽泵，目前国内外没有原型泵可供参考，以及较少运行经验，开展AP1000主泵维修可行性研究对核电厂商运后的安全和经济性具有重要意义。分析了主泵寿命期内故障和失效原因；从现场拆卸，吊装，去污，解体，检修和部件更换以及回装试验等方面分析判断AP1000屏蔽式主泵在60年设计寿命期内维修的必要性和可行性。The nuclear power plant reactor coolant pump is the first class nuclear safety equipment. The operation reliability is directly related to power plant nuclear safety, radiation safety and economy, the operation reliability is also vital to prevent the occurrence and development of nuclear power plants accident. The overhaul of the failure main coolant pump may be one of the important tasks during the overhaul period. AP1000 reactor main pump is a unipolar, centrifugal canned motor pump with flywheel and canned motor. There is no prototype pump and less operation experience for reference at domestic and abroad. The feasibility study of repairing the AP1000 reactor coolant pump is significant for the safety and economy of nuclear power plant after commercial operation. The main pump failure case during the life of pump is analyzed. The maintenance necessity and feasibility about AP1000 canned main pump over the 60-year design life are analyzed from the aspects of removal, lifting, decontamination, disassembling, maintenance, component replacement and assembly test.

Dynamical modeling investigation for economy of nuclear power plants (NPPs) in global nuclear energy market

Non-linear dynamical analysis for the global nuclear energy market is investigated. Currently, the market means a different characteristics comparing to the past situation which had been done before two severe accidents as the Three Mile Island nuclear power plant (NPP) accident in the United States and the Chernobyl NPP disaster in the Soviet Union. For the nuclear related facility, the environmental and safety aspects are the important issues of the analysis. Fundamentally, the economic factor is still a critical matter for the commercial trade between two countries which depend on the energy demand and uranium price. The dynamics simulations show the trend of trade is affected by the several kinds of the aspects. Using system dynamics (SD) method, the event quantification is performed for the event flows, stocks, and feedback where the single and double arrow lines are incorporated.

DEVELOPMENT OF STANDARD COMMUNICATION PROTOCOL FOR EMERGENCY MANAGEMENT OF MAIN CONTROL ROOM OPERATORS IN NUCLEAR POWER PLANTS

In abnormal or emergency situations, correct communication between main control room (MCR) operators is important to safe operation of safety-critical systems such as nuclear power plants (NPPs) and large-scale chemical plants. To improve the communications between MCR operators in NPPs during emergency situations, we developed a standard communication protocol for the emergency operation procedures (EOPs) of Korean Standard Nuclear Plant (KSNP). A general comment from real NPP operators was that the standard communication protocol is expected to increase the safety and economy of NPPs by reducing communication-related human errors.