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.

A Novel Computerized Water Level Control System of PWR Steam Generator of Nuclear Power Plant

This paper presents a novel method to solve old problem of water level control system of pressurized water reactor (PWR) steam generator (SG) of nuclear power plant (NPP) .The level control system of SG plays an important role which effects the reliablity,safty,cost of SG and its mathematical models have been solved.A model of the conventional controller is presented and the existing problems are discussed. A novel rule based realtime control technique is designed with a computerized water level control (CWLC) system for SG of PWR NPP.The performance of this is evaluated for full power reactor operating conditions by applying different transient conditions of SG′s data of Qinshan Nuclear Power Plant (QNPP).

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.

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.

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.

Advanced Coordinating Control System for Power Plant

The coordinating control system is popular used in power plant.This paper describes the advanced coordinating control by control methods and optimal operation,introduces their principals and features by using the examples of power plant op- eration.It is wealthy for automation application in optimal power plant operation.

A framework for nuclear power plant emergency response system

The purpose of this study is to establish an intelligent expert system for nuclear power plant emergency response.A new framework of environmental risk management methodology by the concept of pattern recognition was introduced in this paper.A knowledge-based decision support system for emergency response and risk management of nuclear power plant was also discussed.The mathematical pattern relationship of accidental release effects on neighboring area and the corresponding response measures were presented in this paper.With this decision system,the decision maker can specify the procedure and minimize their human error in the decision process.The improvement of risk response and the quality of management system could be upgraded by this system.Besides,the methodology can also be served as a basis for the future development of environmental risk response system design.

Study on evaluation system for Chinese nuclear power plants

This paper analyzes the meaning, structure, function and assessment methods of a nuclear power plant evaluation system, and the similarities and differences among various assessment methods. Based on this research an integrated and detailed suggestion is proposed on how to establish and improve internal and external evaluation systems for Chinese NPPs. It includes： to prepare and implement the nuclear power plant operational management program, to build an integrated performance indicator system, to improve the present audit system and conduct the comprehensive evaluation system, to set up and implement the integrated corrective action system, to position precisely the status of operation assessment of nuclear power plants, to conduct the assessment aetivilies on constructing NPP, to initiate the specific assessment in some important areas, to establish industrv performance indicator system, to improve the assessment methods, to share the assessment results, to select, cultivate and certify the reviewers, and to enhance interuational communication and cooperation.

Some Technical Solutions for Environmental Protection System during Accidents at Nuclear Power Plants

The paper reports some technical solutions, which suggested or used for increasing of environmental protection during accidents at NPPs. For NNPs with two protective shells and pressure release system such as WWER-1000 a comprehensive, passive-mode environmental protection system of decontamination of the radioactive steam-air mixture from the containment and the intercontainment area was suggested. This system includes the “wet” stage (scrubbers, etc.), the “dry” stage (sorption module), and also an ejector, which in a passive mode is capable of solving the multi-purpose task of decontamination of the air-steam mixture. For WWER-440/230 NPPs three protection levels: 1) a jet-vortex condenser;2) the spray system;3) a sorption module were suggested and installed. For modern designs of new generation NPPs, which do not provide for pressure release systems, a new passive filtering system together with the passive heat-removal system, which can be used during severe accidents in case all power supply units become unavailable, was proposed and after modernization was installed at the KudanKulam NPP (India).

Research on Monitoring and Earlywarning System of Marine Organisms for the Intake of Nuclear PowerPlants

An monitoring and earlywarning system is proposed for marine organisms and the cause of water intake blockage is analyzed. Based onthe intelligent sensing technology, computer software and hardware technology and digital signal processing technology, the buoy monitoring platformsystem is developed by internet of things technology, cloud computing and the application of large data. Remote realtime monitoring of aquatic organisms and foreign bodies is realized based on underwater acoustic detection and low light imaging technology. Data processing center is established to store, analyze and process monitoring information and display it in real time, and provide emergency decision support. Through development and test of relevant key equipments, the reliability of cold source system of nuclear power plants is improved, which effectively reduces theinfluence of marine biological invasion on security and economic operation of the units.

Systematic Approach to Training for Design of Nuclear Power Plant Decommissioning Training in South Korea

In 1979, unavailability of MFWS （Main Feedwater System） in TMI （Three Mile Island） Unit-2 happened in the United States. To make it worse, due to malfunction of Isolation Control Valves in AFWS （Auxiliary Feedwater System）, the supply of cooling water to SGs （Steam Generators） was delayed approximately 8 minutes compared to AOP （Abnormal Operating Procedure）. In the long run, on account of deferred heat sink provision to SGs, the reactor core was melted partially. It was the first critical accident in the US commercial NPP （Nuclear Power Plant） history. Hence, after TMI Accident, US NRC （Nuclear Regulatory Committee） suggested more than one hundred alternatives to improve safety and reliability of NPP. Among these countermeasures, one proposal was related to training area. It was SAT （Systematic Approach to Training） methodology. Therefore, the goal of SAT is the enhancement of NPP stability through training point of view. Since the appearance of SAT in the nuclear industry, it has acquired the unwavering position in the US NPP training field. Meanwhile, significance of NPP decommissioning has been soared up in South Korea since the announcement of Kori Unit-1 decommissioning decision. According to the proclaimed plan from Korean government, Kori Unit-1 is scheduled to be decommissioned from June, 2017. Under this circumstance, nurturing sufficient number of NPP decommissioning engineers is one of the most urgent issues in South Korean nuclear industry. Hence, to elevate efficiency and consistency of training quality, SAT methodology can be a reliable solution for the training of decommissioning engineers. For this reason, the present study aims to design SAT based NPP decommissioning engineer training in South Korea.

A method of parsing based on improving the electricity behavior and regulating excitation system of the diesel generating set in the nuclear power plant

This paper introduces an applicable test plan for emergency diesel generator in nuclear power plant. It advances improvement approaches with problems found during field commissioning test and its trouble-shooting processes. The method is based on the integration of complementary through, the extension theory of matter-element model and neural network theory combine to overcome a neural network to learn shelters, and other defects. The purpose of this paper is to provide the better running and commissioning experience for the similar emergency generator unit.

Integrated Automation of Power Plant Electrical System

Aiming at the comparatively laggard level of power plant electrical system automation, this paper analyzes the feasibility,necessity and some key points of the application of integrated automation technology to power plant electrical system. New idea using fieldbus control system technology is presented. This paper also gives the outline and detailed schemes.

PCTRAN Westinghouse AP1000 Power Control of Pressurized Water Reactor Using Simulink of MATLAB

This paper introduces the simulation, and controls using Simulink of MATLAB for PCTRAN (Personal Computer Transient Analysis) of the power control system (PWR) type pressurized water reactor of PWR WESTINGHOUSE AP1000. The power controller model produces mathematical model description in nonlinear relation form in Simulink of MATLAB which is an important and popular program used at most universities for education. The power controller is described by a block diagram in this paper and some details introduce to clearly understand the work function. The results of action control compared with the PCTRAN programme in modes of automatic and manual control.

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.

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.

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.

Sliding Mode Predictive Control of Main Steam Pressure in Coal-fired Power Plant Boiler

Since the combustion system of coal-fired boiler in thermal power plant is characterized as time varying, strongly coupled, and nonlinear, it is hard to achieve a satisfactory performance by the conventional proportional integral derivative (PID) control scheme. For the characteristics of the main steam pressure in coal-fired power plant boiler, the sliding mode control system with Smith predictive structure is proposed to look for performance and robustness improvement. First, internal model control (IMC) and Smith predictor (SP) is used to deal with the time delay, and sliding mode controller (SMCr) is designed to overcome the model mismatch. Simulation results show the effectiveness of the proposed controller compared with conventional ones.