Multi-objective optimization and evaluation of supercritical CO_(2) Brayton cycle for nuclear power generation

The supercritical CO_(2) Brayton cycle is considered a promising energy conversion system for Generation IV reactors for its simple layout,compact structure,and high cycle efficiency.Mathematical models of four Brayton cycle layouts are developed in this study for different reactors to reduce the cost and increase the thermohydraulic performance of nuclear power generation to promote the commercialization of nuclear energy.Parametric analysis,multi-objective optimizations,and four decision-making methods are applied to obtain each Brayton scheme’s optimal thermohydraulic and economic indexes.Results show that for the same design thermal power scale of reactors,the higher the core’s exit temperature,the better the Brayton cycle’s thermo-economic performance.Among the four-cycle layouts,the recompression cycle(RC)has the best overall performance,followed by the simple recuperation cycle(SR)and the intercooling cycle(IC),and the worst is the reheating cycle(RH).However,RH has the lowest total cost of investment(C_(tot))of$1619.85 million,and IC has the lowest levelized cost of energy(LCOE)of 0.012$/(kWh).The nuclear Brayton cycle system’s overall performance has been improved due to optimization.The performance of the molten salt reactor combined with the intercooling cycle(MSR-IC)scheme has the greatest improvement,with the net output power(W_(net)),thermal efficiencyη_(t),and exergy efficiency(η_(e))improved by 8.58%,8.58%,and 11.21%,respectively.The performance of the lead-cooled fast reactor combined with the simple recuperation cycle scheme was optimized to increase C_(tot) by 27.78%.In comparison,the internal rate of return(IRR)increased by only 7.8%,which is not friendly to investors with limited funds.For the nuclear Brayton cycle,the molten salt reactor combined with the recompression cycle scheme should receive priority,and the gas-cooled fast reactor combined with the reheating cycle scheme should be considered carefully.

Increasing the Efficiency and Level of Environmental Safety of Pro-Environmental City Heat Supply Technologies by Low Power Nuclear Plants

In connection with the current prospect of decarbonization of coal energy through the use of small nuclear power plants (SNPPs) at existing TPPs as heat sources for heat supply to municipal heating networks, there is a technological need to improve heat supply schemes to increase their environmental friendliness and efficiency. The paper proves the feasibility of using the heat-feeding mode of ASHPs for urban heat supply by heating the network water with steam taken from the turbine. The ratio of electric and thermal power of a “nuclear” combined heat and power plant is given. The advantage of using a heat pump, which provides twice as much electrical power with the same heat output, is established. Taking into account that heat in these modes is supplied with different potential, the energy efficiency was used to compare these options. To increase the heat supply capacity, a scheme with the use of a high-pressure heater in the backpressure mode and with the heating of network water with hot steam was proposed. Heat supply from ASHPs is efficient and environmentally friendly even in the case of significant remoteness of heat consumers.

Conceptual Strategy for Mitigating the Risk of Hydrogen as an Internal Hazard in Case of Severe Accidents at Nuclear Power Plant Considering Existing Risks and Uncertainties Associated with the Use of Traditional Strategies

Hydrogen challenge mitigation stands as one of the main objectives in the management of severe accidents at Nuclear Power Plants (NPPs). Key strategies for hydrogen control include atmospheric inertization and hydrogen removal with Passive Autocatalytic Recombiners (PARs) being a commonly accepted approach. However, an examination of PAR operation specificity reveals potential inefficiencies and reliability issues in certain severe accident scenarios. Moreover, during the in-vessel stage of severe accident development, in some severe accident scenarios PARs can unexpectedly become a source of hydrogen detonation. The effectiveness of hydrogen removal systems depends on various factors, including the chosen strategies, severe accident scenarios, reactor building design, and other influencing factors. Consequently, a comprehensive hydrogen mitigation strategy must effectively incorporate a combination of strategies rather than be based on one strategy, taking into consideration the probabilistic risks and uncertainties associated with the implementation of PARs or other traditional methods. In response to these considerations, within the framework of this research it has been suggested a conceptual strategy to mitigate the hydrogen challenge during the in-vessel stage of severe accident development.

Effects of sequential decay on collective flows and nuclear stopping power in heavy-ion collisions at intermediate energies

In this study, the rapidity distribution, collective flows, and nuclear stopping power in ^(197)Au+^(197)Au collisions at intermediate energies were investigated using the ultrarelativistic quantum molecular dynamics(UrQMD) model with GEMINI++ code. The UrQMD model was adopted to simulate the dynamic evolution of heavy-ion collisions, whereas the GEMINI++ code was used to simulate the decay of primary fragments produced by UrQMD. The calculated results were compared with the INDRA and FOPI experimental data. It was found that the rapidity distribution, collective flows, and nuclear stopping power were affected to a certain extent by the decay of primary fragments, especially at lower beam energies. Furthermore, the experimental data of the collective flows and nuclear stopping power at the investigated beam energies were better reproduced when the sequential decay effect was included.

PSA study of the effect of extreme snowfall on a floating nuclear power plant:case study in the Bohai Sea

This study presents a probabilistic safety analysis(PSA)method for the external event of extreme snowfall on a floating nuclear power plant(FNPP)deployed in the Bohai Sea.We utilized the Weibull and Gumbel extreme value distributions to fit the collected meteorological data and obtained a hazard curve for the event of an extreme snowfall where the FNPP is located,providing a basis for the frequency of extreme snowfall-initiating events.Our analysis indicates that extreme snowfall primarily affects the ventilation openings of the equipment,leading to the failure of devices such as the diesel generators.Additionally,extreme snowfall can result in a loss of off-site power(LOOP).Therefore,the developed extreme snowfall PSA model is mainly based on the LOOP event tree,considering responses such as snowfall removal by personnel.Our calculations indicate a core damage frequency(CDF)of 1.13×10^(-10) owing to extreme snowfall,which is relatively low.The results of the cut-set analysis indicate that valve failures in the core makeup tank(CMT),passive residual heat removal system(PRS),and in-containment refueling water storage tank(IRWST)significantly contribute to the CDF.

Development and Research of Non⁃Stirring Conveying Device for Waste Resin in Nuclear Power Plant

Blockage in the storage and transportation of waste resin is a difficult problem in the radioactive waste treatment process of nuclear power plants.In this study,in order to solve the problems of unstable resin transport concentration and easy blockage of conveying equipment and pipelines in nuclear power plants in China,a set of non⁃stirring conveying devices is developed,and theoretical calculations,simulation analysis and experimental verification are carried out.By transporting resin using the no stirring conveying device developed in this paper,it is not only to eliminate the risk of blockage and ensure the safety of transportation,but also to adjust the concentration of conveying resin to change the transport efficiency according to the operating conditions.The effective bearing rate of waste resin storage tank can be improved,so that the comprehensive performance of waste resin storage and transportation in nuclear power plants can be greatly improved.

Assessment of Axial Power Peaking Factors in GHARR-1 LEU Core: A Decadal Simulation Analysis

This study aims to thoroughly investigate the axial power peaking factors (PPF) within the low-enriched uranium (LEU) core of the Ghana Research Reactor-1 (GHARR-1). This study uses advanced simulation tools, like the MCNPX code for analysing neutron behavior and the PARET/ANL code for understanding power variations, to get a clearer picture of the reactor’s performance. The analysis covers the initial six years of GHARR-1’s operation and includes projections for its whole 60-year lifespan. We closely observed the patterns of both the highest and average PPFs at 21 axial nodes, with measurements taken every ten years. The findings of this study reveal important patterns in power distribution within the core, which are essential for improving the safety regulations and fuel management techniques of the reactor. We provide a meticulous approach, extensive data, and an analysis of the findings, highlighting the significance of continuous monitoring and analysis for proactive management of nuclear reactors. The findings of this study not only enhance our comprehension of nuclear reactor safety but also carry significant ramifications for sustainable energy progress in Ghana and the wider global context. Nuclear engineering is essential in tackling global concerns, such as the demand for clean and dependable energy sources. Research on optimising nuclear reactors, particularly in terms of safety and efficiency, is crucial for the ongoing advancement and acceptance of nuclear energy.

基于Power BI的核电大修工单成本分析工具开发与应用

介绍了基于Power BI搭建大修工单成本分析数据模型的开发过程以及应用情况。通过与工单系统的动态关联,准确统计出大修某个工单、某个系统、某个子项、某个区域或是某次大修的成本。基于对大修成本数据的精细分析,从单纯的财务指标计划驱动完成了向业务精益驱动的过程管控转变,持续挖掘大修成本的精益空间,单次大修直接经济效益均超过1000万元,累计效益达7648万元,同时提升成本管理的事前决策支持与业务指导能力,促进大修业务管理水平的持续提升。

SRDAAR－QNPP：a computer code system for the real－time dose assessment of an accident release for Qinshan Nuclear Power Plant

The paper presents a computer code system 'SRDAAR- QNPP' for the real-time dose as-sessment of an accident release for Qinshan Nuclear Power Plant. It includes three parts:thereal-time data acquisition system, assessment computer. and the assessment operating code system. InSRDAAR-QNPP, the wind field of the surface and the lower levels are determined hourly by using amass consistent three-dimension diasnosis model with the topographic following coordinate system.A Lagrangin Puff model under changing meteorological condition is adopted for atmosphericdispersion, the correction for dry and wet depositions. physical decay and partial plume penetrationof the top inversion and the deviation of plume axis caused by complex terrain have been taken in-to account. The calculation domain areas include three square grid areas with the sideline 10 km, 40krn and 160 km and a grid interval 0.5 km, 2.0 km, 8.0 km respectively. Three exposure pathwaysare taken into account:the external exposure from immersion cloud and passing puff, the internalexposure from inhalation and the external exposure from contaminated ground. This system is ableto provide the results of concentration and dose distributions within 10 minutes after the data havebeen inputed.

Study on integrated development and hybrid operation mode of nuclear power plant and pumped-storage power station

The nuclear power plant is suitable for base-load operation, while the pumped-storage unit mainly gives play to capacity benefit in the electric power system;hence, the integrated development and hybrid operation mode of the two can better meet the needs of the electric power system. This article first presents an analysis of the necessity and superiority of such mode, then explains its meaning and analyzes the working routes. Finally, it proposes the business modes as follows: low price pumping water electricity plus nuclear power in the near term;nuclear power shifted to pumped storage power participating in market competition in the middle term;and, in the long term, nuclear power shifted to pumped storage power as primary and serving as an electric power system when needed.

China＇s nuclear power under the global 1.5℃ target： Preliminary feasibility study and prospects

This study explores the measures to achieve the global 1.5 ℃ temperature rise target （1.5 ℃ target） by analyzing the feasibility and obstacles of nuclear power in China. The 1.5 ℃target imposes stricter requirements on China＇s nuclear power. Considering the available nuclear power plant sites, nuclear power layout, equipment manufacture ＆ supply, nuclear power plant construction capacity, supportive operation ＆ management talents, investment, cost effectiveness, and public acceptance, the achievement of the development objectives of nuclear power in China considering the 1.5 ℃ Target is difficult. However, it is possible if favorable decisions and policies are made.

Investigation on Flow Accelerated Corrosion Mitigation for Secondary Circuit Piping of the Third Qinshan Nuclear Power Plant

Flow accelerated corrosion（FAC） is the main failure cause of the secondary circuit carbon steel piping in nuclear power plants.The piping failures caused by FAC have resulted in numerous unplanned outages and tragic fatalities.The existing researches focus on the main factors contributing to FAC,which include metallurgical factors,environmental factors and hydrodynamic factors. Some effective FAC management methods and programs with long term monitoring and inspection data analysis are recommended.But a comprehensive FAC management system should be developed in order to mitigate and manage FAC systematically.In this paper,the FAC influencing factors are analyzed in combination with the operating conditions of the secondary circuit piping in the Third Qinshan Nuclear Power Plant（TQNPP）,China（Third Qinshan Nuclear Power Company Limited,China）.A comprehensive FAC mitigation and management system is developed for TQNPP secondary circuit piping.The system is composed of five processes,viz.materials substitution,water chemical optimization,long-term monitor strategy for the susceptible piping,integrity evaluation of the local thinning defects,and repair or replacement.With the implementation of the five processes,the material of FAC sensitive pipe fittings are modified from carbon steel to stainless steel,N_2H_4 and NH_3 are finally selected as the water chemical regulator of secondary circuit,the secondary circuit pips are classified according to FAC susceptibility in order to conduct long term monitoring strategy,and an integrity evaluation flow for local thinning caused by FAC in carbon steel piping is developed.If the component with local thinning defects is not fit-for-service,corresponding repair or replacement should be conducted.The comprehensive FAC mitigation and management system with five interrelated processes would be a cost-effective method of increasing personnel safety,plant safety and availability.

Safety and effective developing nuclear power to realize green and low-carbon development

This paper analyzes the role of nuclear power of China's energy structure and industry system. Comparing with other renewable energy the nuclear power chain has very low greenhouse gas emission, so it will play more important role in China's low-carbon economy. The paper also discussed the necessity of nuclear power development to achieve emission reduction, energy structure adjustment, nuclear power safety,environmental protection, enhancement of nuclear power technology, nuclear waste treatment, and disposal, as well as nuclear power plant decommissioning. Based on the safety record and situation of the existing power plants in China, the current status of the development of world nuclear power technology, and the features of the independently designed advanced power plants in China, this paper aims to demonstrate the safety of nuclear power. A nuclear power plant will not cause harm either to the environment and nor to the public according to the real data of radioactivity release, which are obtained from an operational nuclear plant. The development of nuclear power technology can enhance the safety of nuclear power. Further, this paper discusses issues related to the nuclear fuel cycle, the treatment, and disposal strategies of nuclear waste, and the decommissioning of a nuclear power plant, all of which are issues of public concern.

Optimization of a dynamic uncertain causality graph for fault diagnosis in nuclear power plant

Fault diagnostics is important for safe operation of nuclear power plants(NPPs). In recent years, data-driven approaches have been proposed and implemented to tackle the problem, e.g., neural networks, fuzzy and neurofuzzy approaches, support vector machine, K-nearest neighbor classifiers and inference methodologies. Among these methods, dynamic uncertain causality graph(DUCG)has been proved effective in many practical cases. However, the causal graph construction behind the DUCG is complicate and, in many cases, results redundant on the symptoms needed to correctly classify the fault. In this paper, we propose a method to simplify causal graph construction in an automatic way. The method consists in transforming the expert knowledge-based DCUG into a fuzzy decision tree(FDT) by extracting from the DUCG a fuzzy rule base that resumes the used symptoms at the basis of the FDT. Genetic algorithm(GA) is, then, used for the optimization of the FDT, by performing a wrapper search around the FDT: the set of symptoms selected during the iterative search are taken as the best set of symptoms for the diagnosis of the faults that can occur in the system. The effectiveness of the approach is shown with respect to a DUCG model initially built to diagnose 23 faults originally using 262 symptoms of Unit-1 in the Ningde NPP of the China Guangdong Nuclear Power Corporation. The results show that the FDT, with GA-optimized symptoms and diagnosis strategy, can drive the construction of DUCG and lower the computational burden without loss of accuracy in diagnosis.

Study on Lightning Protection Design of DCS in a Nuclear Power Plant

DCS （distributed control system） plays a decisive role in the overall operation of a nuclear power plant. If DCS fails, it will seriously affect the normal production of nuclear power plant, causing great losses. So it is very important to take perfect lightning protection measures on DCS of the nuclear power plant. In this paper, according to the actual situation of DCS in a nuclear power plant, by controlling lightning point, securely booting lightning into the ground network, improving low-resistance ground network, eliminating ground loops, determining the safety space, surge protection of power and signal, a set of complete lightning protection design scheme was systematically put forward. Some specific lightning protection measures were highlighted, such as the DCS grounding, equipotential bonds and shields, and some specific considerations were put forward. All of these could offer reference in the practical application.

Fault prediction method for nuclear power machinery based on Bayesian PPCA recurrent neural network model

Early fault warning for nuclear power machinery is conducive to timely troubleshooting and reductions in safety risks and unnecessary costs. This paper presents a novel intelligent fault prediction method, integrated probabilistic principal component analysis(PPCA), multi-resolution wavelet analysis, Bayesian inference, and RNN model for nuclear power machinery that consider data uncertainty and chaotic time series. After denoising the source data, the Bayesian PPCA method is employed for dimensional reduction to obtain a refined data group. A recurrent neural network(RNN) prediction model is constructed, and a Bayesian statistical inference approach is developed to quantitatively assess the prediction reliability of the model. By modeling and analyzing the data collected on the steam turbine and components of a nuclear power plant, the results of the goodness of fit, mean square error distribution, and Bayesian confidence indicate that the proposed RNN model can implement early warning in the fault creep period. The accuracy and reliability of the proposed model are quantitatively verified.

Enhanced graph-based fault diagnostic system for nuclear power plants

Scheduled maintenance and condition-based online monitoring are among the focal points of recent research to enhance nuclear plant safety.One of the most effective ways to monitor plant conditions is by implementing a full-scope,plant-wide fault diagnostic system.However,most of the proposed diagnostic techniques are perceived as unreliable by operators because they lack an explanation module,their implementation is complex,and their decision/inference path is unclear.Graphical formalism has been considered for fault diagnosis because of its clear decision and inference modules,and its ability to display the complex causal relationships between plant variables and reveal the propagation path used for fault localization in complex systems.However,in a graphbased approach,decision-making is slow because of rule explosion.In this paper,we present an enhanced signed directed graph that utilizes qualitative trend evaluation and a granular computing algorithm to improve the decision speed and increase the resolution of the graphical method.We integrate the attribute reduction capability of granular computing with the causal/fault propagation reasoning capability of the signed directed graph and comprehensive rules in a decision table to diagnose faults in a nuclear power plant.Qualitative trend analysis is used to solve the problems of fault diagnostic threshold selection and signed directed graph node state determination.The similarity reasoning and detection ability of the granular computing algorithm ensure a compact decision table and improve the decision result.The performance of the proposed enhanced system was evaluated on selected faults of the Chinese Fuqing 2 nuclear reactor.The proposed method offers improved diagnostic speed and efficient data processing.In addition,the result shows a considerable reduction in false positives,indicating that the method provides a reliable diagnostic system to support further intervention by operators.

Metallographic etching and microstructure characterization of NiCrMoV rotor steels for nuclear power

The grain size of prior austenite has a distinct influence on the microstructure and final mechanical properties of steels. Thus, it is significant to clearly reveal the grain boundaries and therefore to precisely characterize the grain size of prior austenite. For NiCrMoV rotor steels quenched and tempered at high temperature, it is really difficult to display the grain boundaries of prior austenite clearly, which limits a further study on the correlation between the properties and the corresponding microstructure. In this paper, an effective etchant was put forward and further optimized. Experimental results indicated that this agent was effective to show the details of grain boundaries, which help analyze fatigue crack details along the propagation path. The optimized corrosion agent is successful to observe the microstructure characteristics and expected to help analyze the effect of microstructure for a further study on the mechanical properties of NiCrMoV rotor steels used in the field of nuclear power.

Transient Analysis of Steam Generator in PWR Nuclear Power Plant

The water level control system of steam generator in a pressurized water reactor of nuchear power plant plays an important role which effects the water level control of the steam generator are due the reverse dynamics behavior,so the transient analysis of the steam generator should firstly solve their mathematical models.For determination of dynamic behavior and design and testing of the control system, a nonlinear math model is developed using one dimensional conservation equations of mass,momentum and energy of primary and secondary sides of the steam generator. The nonlinear model is verified with standard power plant data available in the references, then the steady states and transient calculations are performed for full power to 5% power reactor operation of the steam generator of Chinese Qinshan Nuclear Power Plant.

Development and Application of Maintenance Template in Pressurized Water Reactor Nuclear Power Plant

Good practices of maintenance optimization in nuclear power field need to be effectively consolidated and inherited,and maintenance optimization can provide technology support to create a long-term reliable and economic operation for nuclear power plants( NPPs) especially for a large number of nuclear powers under construction. Based on the development and application of maintenance template in developed countries,and combining with reliability-centered maintenance( RCM) analysis results and maintenance experience data over the past ten years in domestic NPPs, the development process of maintenance template was presented for Chinese pressurized water reactor( PWR) NPP,and the application of maintenance template to maintenance program development and maintenance optimization combined with cases were demonstrated. A shortcut was provided for improving the efficiency of maintenance optimization in domestic PWR NPP,and help to realize a safe,reliable,and economic operation for domestic NPPs.