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.

Research on Verification Method of Motor Startups in Nuclear Power Plants Based on Topology Recognition

There are many motors in operation or on standby in nuclear power plants,and the startup of group motors will have a great impact on the voltage of the emergency bus.At present,there is no special or inexpensive software to solve this problem,and the experience of engineers is not accurate enough.Therefore,this paper developed a method and system for the startup calculation of group motors in nuclear power plants and proposed an automatic generation method of circuit topology in nuclear power plants.Each component in the topology was given its unique number,and the component class could be constructed according to its type and upper and lower connections.The subordination and topology relationship of switches,buses,and motors could be quickly generated by the program according to the component class,and the simplified direct power flow algorithm was used to calculate the power flow for the startup of group motors according to the above relationship.Then,whether the bus voltage is in the safe range and whether the voltage exceeds the limit during the startup of the group motor could be judged.The practical example was used to verify the effectiveness of the method.Compared with other professional software,the method has high efficiency and low cost.

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.

The Information Security of Digital Control Systems for Nuclear Power Plants

The NPP （nuclear power plant） being vital objects of an energy infrastructure must be protected against malicious actions affecting their safety, and cyber security plays a key part in attaining this goal. The paper considers, implemented by the authors within the project of advanced digital control system for NPP with the reactor VVER-1000, a system of unauthorized access protection, partially built up on the technology of AA （active audit） and expert system. The AA technology is based on response of the system on deviation of current signature of the automated process control system from stable state rather than on a certain signature of attack and relies on the estimation of the behavioral models of the particular digital control system. The advent of active audit reflects the current situation in the digital control systems where complex distributed platforms are used to construct automated process control system. The active audit allows one to make the digital control system functionally closed, provided that it is determinate. The methodology of the active audit does not give u external （barrier） and traditional （password, antivirus） methods of unauthorized access protection. These methods can be used when it is appropriate to achieve a required protection level.

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.

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.

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.

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.

Development of SA-533 Type B CL. 1+SA-240 Type 304L roll-bonded clad steel plate for safety injection tank of CAP1400 nuclear power plant

Aiming to meet the demand of the country＇ s nuclear demonstration project on the CAP1400 nuclear power plant, Baosteel uses the roll-bonding technology and develops the SA-533 Type B CL. 1 ＋ SA-240 Type 304L high-strength and high-toughness clad steel plate with a shear strength of over 310 MPa for the nuclear power plant＇ s safety injection tank. The properties of the quenched and tempered and the simulated post-weld heat treatment states are systematically studied herein through a comprehensive inspection and evaluation of the composition,microstructure,and properties of the clad steel plate. The results show that the bonding interface has high shear strength and that the base metal has high strength and good toughness at low temperatures. Hence, the performance fully meets the technical requirements of the CAP1400 nuclear power plant＇ s safety injection tank in the country＇ s nuclear demonstration project. The roll-bonded clad steel plate can be used to manufacture the safety injection tank of the CAP1400 nuclear power plant.

Probabilistic Approach of Coastal Defense Against Typhoon Attacks for Nuclear Power Plant

With the global warming and sea level rising, it is widely recognized that there is an increasing tendency of typhoon occurrence frequency and intensity. The defenses code against typhoon attacks for nuclear power plant should be calibrated because of the increasing threat of typhoon disaster and severe consequences. This paper discusses the probabilistic approach of definitions about ＂probable maximum typhoon＂ and ＂probable maximum storm surge＂ in nuclear safety regulations of China and has made some design code calibrations by use of a newly proposed Double Layer Nested Mtdti-objective Probability Model （DLNMPM）.

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.

Aseismic safety analysis of a prestressed concrete containment vessel for CPR1000 nuclear power plant

Abstract： The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete containment vessel （PCCV） model with a semi-infinite elastic foundation and practical arrangement of tendons has been established to analyze the aseismic ability of the CPR1000 PCCV structure under seismic loads and internal pressure. A method to model the prestressing tendon and its interaction with concrete was proposed and the axial force of the prestressing tendons showed that the simulation was reasonable and accurate. The numerical results show that for the concrete structure, the location of the cylinder wall bottom around the equipment hatch and near the ring beam are critical locations with large principal stress. The concrete cracks occurred at the bottom of the PCCV cylinder wall under the peak earthquake motion of 0.50 g, however the PCCV was still basically in an elastic state. Furthermore, the concrete cracks occurred around the equipment hatch under the design internal pressure of 0.4MPa, but the steel liner was still in the elastic stage and its leak-proof function soundness was verified. The results provide the basis for analysis and design of containment vessels.

Comparison of Risk Assessment for a Nuclear Power Plant Construction Project Based on Analytic Hierarchy Process and Fuzzy Analytic Hierarchy Process

Recently, plant construction throughout the world, including nuclear power plant construction, has grown significantly. The scale of Korea’s nuclear power plant construction in particular, has increased gradually since it won a contract for a nuclear power plant construction project in the United Arab Emirates in 2009. However, time and monetary resources have been lost in some nuclear power plant construction sites due to lack of risk management ability. The need to prevent losses at nuclear power plant construction sites has become more urgent because it demands professional skills and large-scale resources. Therefore, in this study, the Analytic Hierarchy Process (AHP) and Fuzzy Analytic Hierarchy Process (FAHP) were applied in order to make comparisons between decision-making methods, to assess the potential risks at nuclear power plant construction sites. To suggest the appropriate choice between two decision-making methods, a survey was carried out. From the results, the importance and the priority of 24 risk factors, classified by process, cost, safety, and quality, were analyzed. The FAHP was identified as a suitable method for risk assessment of nuclear power plant construction, compared with risk assessment using the AHP. These risk factors will be able to serve as baseline data for risk management in nuclear power plant construction projects.

Decomposition of oil cleaning agents from nuclear power plants by supercritical water oxidation

Oil cleaning agents generated from nuclear power plants(NPPs)are radioactive organic liquid wastes.To date,because there are no satisfactory industrial treatment measures,these wastes can only be stored for a long time.In this work,the optimization for the supercritical water oxidation(SCWO)of the spent organic solvent was investigated.The main process parameters of DURSET(oil cleaning agent)SCWO,such as temperature,reaction time,and excess oxygen coefficient,were optimized using response surface methodology,and a quadratic polynomial model was obtained.The determination coefficient(R^(2))of the model is 0.9812,indicating that the model is reliable.The optimized process conditions were at 515 C,66 s,and an excess oxygen coefficient of 211%.Under these conditions,the chemical oxygen demand removal of organic matter could reach 99.5%.The temperature was found to be the main factor affecting the SCWO process.Ketones and benzene-based compounds may be the main intermediates in DURSET SCWO.This work provides basic data for the industrialization of the degradation of spent organic solvents from NPP using SCWO technology.

Nuclear power plant fault diagnosis based on genetic-RBF neural network

It is necessary to develop an automatic fault diagnosis system to avoid a possible nuclear disaster caused by an inaccurate fault diagnosis in the nuclear power plant by the operator. Because Radial Basis Function Neural Network (RBFNN) has the characteristics of optimal approximation and global approximation. The mixed coding of binary system and decimal system is introduced to the structure and parameters of RBFNN, which is trained in course of the genetic optimization. Finally, a fault diagnosis system according to the frequent faults in condensation and feed water system of nuclear power plant is set up. As a result, Genetic-RBF Neural Network (GRBFNN) makes the neural network smaller in size and higher in generalization ability. The diagnosis speed and accuracy are also improved.

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.

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.

Development of nuclear power plant real-time engineering simulator

A nuclear power plant real-time engineering simulator was developed based on general-purpose ther- mal-hydraulic system simulation code RELAP5. It mainly consists of three parts:improved thermal-hydraulic system simulation code RELAP5,control and protection system and human-machine interface. A normal transient of CHASHMA nuclear power plant turbine step load change from 100% to 90% of full power,was simulated by the en- gineering simulator as an application example. This paper presents structure and main features of the engineering simulator,and application results are shown and discussed.

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.

The Role of Countermeasures in Mitigating the Radiological Consequences of Nuclear Power Plant Accidents

The countermeasures are the actions that should be taken, after the occurrence of a nuclear accident to protect the public against the associated risks. These actions may be represented by sheltering, evacuation, distribution of stable iodine tablets and/or relocation. This study represents a comprehensive probabilistic study to investigate the role of the adoption of the countermeasures in case of a hypothetical accident of type LOCA for nuclear power plant of PWR （1000 Mw）. The effective doses in different organs, short and long health effects, and the associated risks are calculated with and without countermeasures. In addition, the overall costs of the accident and the costs of countermeasures are estimated which represent our first trials to know how much the proposed accident cost. The results showed that, the area around the site requires early and late countermeasures action after the accident especially in the downwind sectors. For late countermeasures, the duration time of relocation ranged from about two to 10 years. The adoption of the countermeasures increases the costs of emergency plan by 40% but reduces the risk associated the accident.