Generation Of Nuclear Power Plants Research Articles

Evaluation of the vibration response of third generation nuclear power plants with isolation technology under large commercial aircraft impact

With the development of nuclear power technology, third-generation nuclear power plants (NPPs) are being built in many places. Examples include the American passive AP1000 NPP, the French EPR NPP, and China's Hualong One NPP. These will become the main reactor type for future NPP construction. Base isolation is employed as the aseismic technology in NPPs, and its design standard has been standardized according to ASCE 4–16 and other related codes. Since the 9/11 terrorist incident, large commercial aircraft impact (AI), as a beyond design-basis event, has attracted increasing attention in nuclear research. In the case of a third-generation NPP with isolation technology, it is necessary to evaluate its safety and integrity against AI. In this study, the dynamic response of a base-isolated third-generation NPP subjected to impact by a large commercial aircraft, the Airbus A340-300, is investigated. The most typical design for a third-generation NPP is to build many buildings on the same base-mat. The presence of a flexible isolation layer renders the dynamic behavior of a base-isolated NPP different from that of a non-isolated NPP under AI loads. Five key factors, including the horizontal stiffness of isolation bearings, horizontal damping of isolation bearings, total weight and weight distribution of the NPP, and AI force and direction affect the vibration characteristics of NPPs. The horizontal stiffness of isolation bearings affects the constraint effect of the foundation on the superstructure, while the damping of isolation bearings affects the dissipation of impact energy. Meanwhile, the total weight and weight distribution of the NPP affect the vibration response of different regions, and the impact force affects the vibration intensity of the NPP.

Structural safety reliability of concrete buildings of HTR-PM in accidental double-ended break of hot gas ducts

Abstract Safety analysis of nuclear power plant (NPP) especially in accident conditions is a basic and necessary issue for applications and commercialization of reactors. Many previous researches and development works have been conducted. However, most achievements focused on the safety reliability of primary pressure system vessels. Few literatures studied the structural safety of huge concrete structures surrounding primary pressure system, especially for the fourth generation NPP which allows existing of through cracks. In this paper, structural safety reliability of concrete structures of HTR-PM in accidental double-ended break of hot gas ducts was studied by Exceedance Probability Method. It was calculated by Monte Carlo approaches applying numerical simulations by Abaqus. Damage parameters were proposed and used to define the property of concrete, which can perfectly describe the crack state of concrete structures. Calculation results indicated that functional failure determined by deterministic safety analysis was decided by the crack resistance capability of containment buildings, whereas the bearing capacity of concrete structures possess a high safety margin. The failure probability of concrete structures during an accident of double-ended break of hot gas ducts will be 31.18%. Adding the consideration the contingency occurrence probability of the accident, probability of functional failure is sufficiently low.

Fully developed turbulent convection of Lead Bismuth Eutectic in the elementary cell of the NACIE-UP Fuel Pin Bundle

One of the requirements for achieving high levels of safety in fourth generation nuclear power plants, is that core thermal-hydraulics can be simulated numerically at a good level of accuracy. To this aim, detailed validation of turbulent heat transfer models needs to be carried out and discussed, in relevant flow and heat transfer configurations, and for low Prandtl-number fluids.The focus of this research is the turbulent, fully developed convection in a heated rod bundle with a triangular arrangement, and a pitch-to-diameter ratio P/D=1.4. Statistics derived from a set of Direct Numerical Simulations (DNS) at the moderate Reynolds number of Re=8290 are compared to Reynolds Averaged Navier-Stokes (RANS) solutions where the closure is provided by the two-equation model SST k-ω and by the Reynolds-Stress Model. Comparisons are drawn for forced flow and in mixed convection conditions (Ri≈0.25). Global quantities extracted from experiments performed in very similar conditions are also compared against the numerical results. Profiles of the turbulent Prandtl number about the unit flow cell are also displayed and discussed.This work clarifies through detailed comparison that if, on the one hand, only the Reynolds-Stress model is able to reasonably capture important flow features like secondary flow components, on the other hand also the SST k-ω two-equation model considered is acceptably accurate in predicting the integral quantities of interest.

Sensitivity analysis based on Morris method of passive system performance under ocean conditions

Passive safety systems, which do not require external input to operate, are widely used to enhance the inherent safety in new generation nuclear power plants. A challenging issue still to be resolved is the quantification of the functional failure probability of passive safety systems. The evaluation of functional failure relies on repeated thermal-hydraulics simulations which is usually time-consuming in practice. Response surface method has been proposed to tackle this issue. However, the number of input parameter limits its efficiency to a great extent. Therefore, it is necessary to identify the uncertain input parameters that have an important impact on the system performance, which helps reduce the dimension of the input parameters. Especially in marine nuclear power plants, the complex ocean motions will introduce more uncertainties in the passive system behavior. It’s of great significance to screen key parameters for passive safety systems under ocean conditions. In this paper, the Morris method is used to perform an efficient sensitivity analysis to identify key parameters for a passive residual heat removal system in IPWR200. Twenty-four parameters related to passive system behavior are screened and ranked for their sensitivity. The results demonstrated that four of the twenty-four parameters have a more significant influence on the passive system performance.

Fatigue model of domestic 316LN steel in simulated primary coolant environment of CAP1400

Fatigue tests of domestic 316LN were carried out in the simulated primary coolant environment of CAP1400 (the advanced 3rd generation nuclear power plants developed in China). Effects of temperature, strain rate and dissolved oxygen (DO) on fatigue life were analyzed. Based on the results, a fatigue model was developed. Fatigue life in primary coolant environment increases logarithmically as the temperature decreases from 321 ℃ to 150 ℃. When the temperature decreases to 150 ℃, fatigue life in primary coolant environment is very close to that in air environment. Fatigue life increases as the strain rate increases from 0.004%/s to 1%/s. After the strain increases to more than 0.4%/s, fatigue life in primary coolant environment is very close to that in air environment. Effect of DO on fatigue life in primary coolant environment is not obvious in the DO range of 0.005 ppm to 8 ppm. Compared with NUREG6909 model, the present fatigue model is more accurate in lifetime prediction of domestic 316LN.

Combined cycles-coupled high-temperature and very high-temperature gas-cooled reactors: Part II – Engineering design

Next generation nuclear power plant concepts, such as the high-temperature and very high-temperature gas-cooled reactors, have significant potential for addressing ever growing energy needs, but their implementation must first overcome several engineering limitations. Accordingly, based on the theoretical optimization performed in the Part I of this study, the parameters of combined cycle-coupled high-temperature and very high-temperature gas-cooled reactors were evaluated with the consideration of current engineering limitations including the compressor pressure ratio, reactor inlet temperature, turbine blade cooling, main steam parameters, and exhaust humidity of the steam turbine. Four different combined cycles were then designed and compared with respect to these five parameters. The resulting design values were also compared to the theoretical optimization values in Part I of this paper, revealing that the ratio of the design value to the theoretical value decreased with increasing reactor outlet temperature. The comparison also suggested that the combined cycle-coupled very high-temperature gas-cooled reactor has significant potential, as its cycle efficiency could be further improved once the identified engineering restrictions are overcome.

Using next generation nuclear power reactors for development of a techno‐economic model for hydrogen production

Nuclear and hydrogen are considered to be the most promising alternatives energy sources in terms of meeting future demand and providing a CO₂-free environment, and interest in the development of more cost-effective hydrogen production plants is increasing—and nuclear-powered hydrogen generation plants may be a viable alternative. This paper is a report on investigating the application of new generation nuclear power plants to hydrogen production and development of an associated techno-economic model. In this paper, theoretical and computational assessments of generations II, III+, and IV nuclear power plants for hydrogen generation scenarios have been reported. Technical analyses were conducted on each reactor type—in terms of the design standard, fuel specification, overnight capital cost, and hydrogen generation. In addition, a theoretical model was developed for calculating various hydrogen generation parameters, and it was then extended to include an economic assessment of nuclear power plant-based hydrogen generation. The Hydrogen Economic Evaluation Program originally developed by the International Atomic Energy Agency was used for calculating various parameters, including hydrogen production and storage costs, as well as equity, operation and maintenance (O&M), and capital costs. The results from each nuclear reactor type were compared against reactor parameters, and the ideal candidate reactor was identified. The simulation results also verified theoretically proven results. The main objective of the research was to conduct a prequalification assessment for a cogeneration plant, by developing a model that could be used for technical and economic analysis of nuclear hydrogen plant options. It was assessed that high-temperature gas-cooled reactors (HTGR-PM and PBR200) represented the most economical and viable plant options for hydrogen production. This research has helped identify the way forward for the development of a commercially viable, nuclear power-driven, hydrogen generation plant.

A smart component methodology for reliability analysis of dynamic systems

The new generation of nuclear power plants are being designed with passive safety systems and advanced digital Instrumentation & Control (I&C) systems for achieving required operational performance and improved reliability. Digital I&C systems often perform complex tasks while interacting with time dependent process dynamics. Similarly, in passive decay heat removal systems, the time dependence arising from the variation of heat load and interaction between redundant loops are important. Since static reliability methods cannot model these systems without significant approximations, dynamic reliability methods are preferred. Dynamic methods improve the correctness of modeling and shift the burden of proof for correctness from the modeling expert to the methodology. Though a number of dynamic reliability methods have been developed, they have not been adopted widely, like that of the traditional event tree/fault tree methods. A significant reason is due to the lack of an intuitive simulation framework to translate and represent the system structure and functional aspects by a faithful system reliability model. Therefore, we propose in this paper a methodology known as Smart Component Methodology (SCM) for reliability modeling of dynamic safety systems. The methodology is based on an intuitive object oriented framework for the representation of the components, system structure, behavior and reliability data of a safety system. A suitable Monte Carlo simulation algorithm is embedded into this framework to quantify the system reliability. The SCM architecture for dynamic system representation and the general Monte Carlo simulation algorithm used for driving the object oriented framework are presented. The method is applied to example I&C systems, passive heat removal systems involving human actions. The results are compared with traditional methods to validate the results and to demonstrate the improved accuracy and ease and generality of modeling.

How fusion power can contribute to a fully decarbonized European power mix after 2050

In the second half of this century, the European energy mix will be very likely completely decarbonized. Two main options are available to generate carbon free electricity: either to rely on renewable energy sources only, or to integrate renewables with a new generation of nuclear power plant, e.g. fusion power plant. In the former case a large storage capacity and/or dispatchable generation are required to compensate for the variable electricity generation. The size of the necessary storage systems and dispatchable power plants can be reduced if a base-load carbon free power technology is available, such as fusion. These options are analysed in this paper with the CO.ME.S.E. (COsto MEdio del Sistema Elettrico – Electric System Average Cost) code that assesses the hourly balance between load and generation, through a simplified dispatch model and the system costs, through a stochastic economic analysis. Two sets of power generation scenarios are considered in typical North and South-Europe cases. The study gives key indications to make fusion a cost-effective ingredient of a future European decarbonized power mix.

Role of Sensors (Nano) in Nuclear Technology

Sensors in a nuclear plant required to be highly accurate, could sense and transfer the process variables at high speeds, provide timely display, Sensors relate with nuclear plant’s processes to measure several parameters such as gas leaks, high temperatures, pressures and flow levels, monitor radiation, chemical and biological nuclear sensing, concentrations of gases and components of safety that process the data enabling on-site/on-line monitoring. Special instruments also measure vibrations, H2 concentration, and conductivity of water, meteorological and concentration or other applications. For example, resistance temperature detectors those are significant to the safety system of the plants[1], with predictable accuracy and respond to changes in temperature in less than 3 seconds. Emerging sensors for the core plant technology has continued the same from their commencement, though several novel sensor technologies have been developed and prototyped, could find adoption in the next- generation nuclear power plants. The next-generation reactors may require extremely high temperatures accurate measurement and this could be the technical challenge that was facing today’s novel sensor development. In this presentation, the author discusses sensors, types, miniaturisation towards nanosensors, why we need to measure, their characteristics, classification with a specific set of applications followed by summary and conclusions. Combining these smart nanosensors with virtual reality / AI with available systems based on nanostructured electronics wireless sensor networks could enable more and more effective functionalities.

Long-term decay heat removal in a submerged SMR

Following the Fukushima-Daiichi nuclear accident in March 2011, innovations are needed to improve the reliability of new generation nuclear power plants toward scenarios where electrical power and ultimate heat sink are lost. This paper describes an integral design and basic passive safety strategy of a Small Modular Reactor (SMR) submerged in the sea or in an artificial lake, then performing a preliminary analysis of the long-term decay heat removal. The analysis considers a pressurized reactor placed in a horizontal cylindrical hull, which is surrounded by the external water. The simulated system is based on the Flexblue concept, developed by French company DCNS (now Naval Group). The object of the investigation is the natural circulation in the submerged containment, which is the key component for the long-term cooling. Following a rupture in the primary circuit, decay heat must be removed according to a fully passive safety strategy for an indefinitely long period. The purpose of this work is to study the effectiveness of a sump natural circulation flow to cool the fuel rods, up to several days after the scram. Decay heat generates steam in the core, which is released in the containment and condensed on the metal surface, transferring the heat to the exterior. Relap5-Mod3.3 has been employed to simulate the accident scenario. Results show the consistency of the safety principles and stimulate experimental investigations. However, the sensitivity analysis identifies the nodalization of the reactor containment as a modeling and numerical issue, deserving further analyses.

Titanium Influence on the Microstructure of FeCrAl Alloys Used for 4R Generation Nuclear Power Plants

4R generation nuclear power plants should work with metallic material of the highest quality, capable to resist in maximum safe conditions for 25-30 years. FeCrAl alloys are capable of such performance, because of the resistance to: oxidation at high temperatures, corrosion, erosion and penetrating radiations in liquid metal environments. In addition, such materials are capable of forming on their surface some oxide layers, textured and self-renewable, with high adhesion to metallic substrates. These properties can be improved by microalloying with metal elements such as Ti, Zr, Y, Hf in amounts of 1 to 3%. These chemical elements have high affinity to oxygen, being able to stabilize the structure of the superficial layer of oxide (alumina) and to increase adhesion to the metallic substrate. The FeCrAl alloys microalloyed with titanium were obtained in a VAR (Vacuum Arc Remelting) equipment in argon atmosphere (99.99% purity). There were obtained three batches microalloyed with 0.5%, 1% and 1.5% titanium, preserving the same Fe-14Cr-5Al metal matrix. In order to determine the chemical composition of the oxide layer and of the sample bulk, the EDAX analysis was performed. Microstructural features were revealed using SEM analysis. The results showed the capacity of the FeCrAl alloy to form oxide layers, with different textures and rich in elements such as Al and Ti. The compositional analysis performed on FeCrAl samples microalloyed with 0.5, 1.0 and 1.5% Ti in the central zone shows a relatively similar composition compared to the technological calculations made, reflecting the homogeneity of the alloy. The microhardness measurements performed on the cross sections of the metallic samples attest values in the range 163-183 HV0.2, falling within the normal range for these materials. These values are influenced by the presence of the alloying elements in the metallic matrix and by the homogeneous arrangement of constituents.

Effects of toolpath and clamping strategies in machining distortion of stainless-steel parts

Heat exchangers in new nuclear power generation plants are made of thin AISI 316L stainless-steel plates stacked together in order to improve their efficiency and compactness. To ensure the assembly, the global distortion of those plates must be mastered and minimized, mainly by predicting the evolution of the residual stress field during their manufacturing process chain. During machining, those residual stresses are redistributed to reach another equilibrium state, leading to a macroscopic part distortion. The main objective of this work is to study experimentally the influence of the machining toolpath and clamping strategies on the global part distortion. Then, a part distortion model is developed in order to verify the clamping effect on the distortion.

Research on High Reliability Power Supply Design Scheme of Nuclear Power Plant

After the Fukushima nuclear accident, the reliability requirements for Nuclear Power Plant (NPP) safety systems have been further improved worldwide. Therefore, it is necessary to provide a safe, reliable and economical scheme of the power supply system to cope with the abnormal conditions. Based on the reliability of the power supply of the 3rd generations of NPPs and combined with the application of the defend in depth concept in the electrical system, this paper provides a brief introduction of the typical 3rd generation NPP electrical system in the following area: the configuration of the electrical power system, defence in depth principle of the power supply, the basic structure of electrical power system. On this basis, an optimal power supply scheme is proposed.

Life Cycle Assessment of the New Generation GT-MHR Nuclear Power Plant

This study describes a life cycle assessment (LCA) of a fourth generation (4G) nuclear power plant. A high temperature helium cooled reactor and gas turbine technology with modular helium reactor (GT-MHR) is used in this study as an example. This is currently one the safest design of a nuclear power plant. The study also takes into account impact of accidents and incidents (AI) which happened around the world at nuclear power generation facilities. The adopted method for the study is a hybrid LCA analysis. The analysis of each phase of the life cycle was done on the basis of process chain analysis (PCA). Where detailed data were not available, the Input/Output (I/O) databases was employed. The obtained results show that greenhouse gases (GHG) emissions and energy intensity per unit of electricity production are relatively low. In fact, these are even lower than emissions from a number of renewable energy sources. The results show considerably different greenhouse gases (GHG) emissions and energy intensity per unit of electricity production when effects of AI are taken into account.

The effect of cooling rates on carbide precipitate and microstructure of 9CR-1MO oxide dispersion strengthened(ODS) steel

Abstract The 9Cr-1Mo ferritic-martensitic ODS steel is a promising structural material for the next generation nuclear power plants including fast reactors for application in reactor vessels and nuclear fuel. The ODS steel was cooled down by furnace cooling, air cooling, oil quenching and water quenching, respectively, after normalizing it at 1150 °C for 1 h and then tempering at 780 °C for 1 h. It is found that grain size, a relative portion of ferrite and martensite, martensitic lath configuration, behaviors of carbide precipitates, and hardness of the ODS steel are strongly dependent on a cooling rate. The grain size and martensitic lath width become smaller with the increase in a cooling rate. The carbides were precipitated at the grain boundaries formed between the ferrite and martensite phases and at the martensitic lath interfaces. In addition, the carbide precipitates become smaller and more widely dispersed with the increase in a cooling rate, resulting in that the faster cooling rate generated the higher hardness of the ODS steel.

Response of growth and development of the Pacific oyster (Crassostrea gigas) to thermal discharge from a nuclear power plant

BackgroundDuring electricity generation of nuclear power plant, heat energy cannot be completely converted into electrical energy, and a part of it is lost in the form of thermal discharge into the environment. The thermal discharge is harmful to flora and fauna leading to environmental deterioration, biological diversity decline, and even biological extinction.ResultsThe present study investigated the influence of thermal discharge from a nuclear power plant on the growth and development of Pacific oyster Crassostrea gigas which is widely used as bio indicator to monitor environmental changes. The growth of soft part and the gonad development of oysters were inhibited due to thermal discharge. During winter season, temperature elevation caused by thermal discharge promoted the growth of oyster shells. During summer season, the growth rate of oysters in thermal discharge area was significantly lower than that of the natural sea area.ConclusionsThe results of this study provided a better understanding of assessing the impact of thermal discharge on the marine ecological environment and mariculture industry. It also provided a scientific basis for defining a safe zone for aquaculture in the vicinity of nuclear power plants.

Study on mechanical behavior of containment in nuclear power plant during prestressing construction

Abstract Taking a third generation nuclear power plant as the research background, a refined model of a prestressed concrete containment which consisted of 270 complex spatial prestressed tendons was accurately established, using FE software ABAQUS. The cooling method was adopted to simulate the tensioning process of tendons. The simulation method of the tensioning of tendons during the prestressing construction of containment was verified by comparing with the measured data on site of the practical project. The mechanical behavior of the containment during the prestressing construction according to the prestressing construction scheme on site was studied in detail. Then the influence of steel liner and reinforcement on mechanical behavior of the containment during the prestressing construction of prestressed tendons was discussed. Finally, four prestressing construction schemes were analyzed and a reasonable optimization scheme was obtained.

Experimental investigation of the isothermal flow field across slant 5-tube bundles in helically coiled steam generator geometry using PIV

Abstract Next generation nuclear power plant designs have been considering Helically Coiled Steam Generators (HCSGs) because of its compactness and capability to absorb thermal expansion. During the design phase, Computational Fluid Dynamics (CFD) has been becoming a powerful tool and being applied to HCSG to predict important engineering parameters including pressure drop and flow induced vibrations. However, the complex geometry of the HCSG requires experimental data to validate the simulation results generated by the highly turbulent flow. In this study, an experimental test facility was constructed with five-rod bundles that coils against adjacent rod to produce high-resolution flow field data in space and time. A Particle Image Velocimetry (PIV) visualized the flow field between adjacent rods in three regions at approximately Re 3600 with up to 5000 frames per second (5 kHz). The flow interaction between the rods and the vortical structure were observed from side view of the flow channel. The slant rod bundle created different flow bifurcation patterns. In a small flow channel between adjacent rods at top and bottom, there exist two opposite directional horizontal flows changing magnitude in time. However, once the pattern was formed, it did not change the pattern but only the magnitude during the current test. The current study more focused on the local flow measurement for a short time period. In further works, relatively large areas within the test section need to be monitored for a longer period to investigate whether the bifurcation pattern changes or not.

A study on tensile properties of Alloy 709 at various temperatures

In recent years, there have been several advancements in energy production from both fossil fuels and the alternate “clean” sources such as nuclear fission. These advancements are fueled by the need for more efficient systems that will optimize the use of the depleting fossil fuel reserves and shift the focus to cleaner sources of energy. The efficiency of any power generation cycle is dependent on the ability of the structural material to withstand increased peak operating temperatures. Advanced austenitic stainless steels have been in the focus as structural material for the next generation nuclear power plants, due to their strength, corrosion resistance, weldability and the wide range of temperatures at which the austenite phase is stable. Alloy 709, a recently developed advanced austenitic stainless steel, is being investigated in this paper. In this study, tensile tests were conducted on dog-bone samples of Alloy 709 in an in-situ scanning electron microscope (SEM) loading and heating stage, equipped with electron backscatter diffraction (EBSD), at various temperatures. The in-situ experiments indicated that the material primarily accommodated deformation by slip at lower temperatures. Void formation and coalescence at grain boundaries preceded slip at higher temperatures. Although crack initiation at all elevated temperatures was intergranular, the crack propagation through the material and the final fracture was transgranular ductile. Additionally, tensile tests were conducted on larger cylindrical samples at 550, 650 and 750 °C in air. The results of tests conducted in air and in-situ were found to be in agreement, at these temperatures.