With the continuous development of the nuclear power technology in the world, all countries in the world are becoming more and more interested in the inherent safety of nuclear power technology, while the research and development of the spherical bed type high temperature gas cooled reactor nuclear power technology in China has formally catered to this demand. As a major national science and technology project, since the construction of the high temperature gas cooled reactor demonstration project (HTR-PM) since 2012, the civil construction of the nuclear island has been basically completed, the installation of equipment has been carried out orderly, and many process systems have entered debugging and operation stage gradually. As an important auxiliary process system, fuel handling and storage system for online refueling of the pebble bed high temperature gas cooled reactor, plays an important role in relation to the stable operation of the reactor. The main functions of the fuel handling and storage system are loading the fresh fuel elements and unloading the spent fuel elements which has reached its target burnup continuously for reactor operation, the spent fuel elements would be discharged into the spent fuel canister firstly, when the spent fuel storage canister is full of spent fuel, the canister would be sealed through welding method, and then the spent fuel canister would be transferred and stored in the spent fuel storage silo with the ground crane system. The fuel element of the pebble bed high temperature gas cooled reactor is spherical fuel element with graphite matrix, the fuel elements will have friction and collision with the inner wall of the pipeline in transporting process, which will produce graphite dust, the graphite dust should be removed continuously though filtration method, so as not to affect the fuel elements transportation in pipeline. This article focus on the production mechanism and filtering method of the graphite dust in graphite matrix fuel element transporting process in pipeline, to study the graphite dust removal technology, and then we could provide theoretical guidance for the design and operation of the key system and equipment for HTR-PM.

The vibration of pipeline system induced by pump is one of the most important factors affecting the safety and stability of nuclear power plant (NPP) operation, and is also the most prominent vibration and noise transmission channel in the plant area. Analysis results show that various support structures such as pipe clamps, hangers and brackets are the main transmission routes of pipeline vibration. However, it is usually difficult to obtain a satisfactory absorption effect by using traditional passive vibration isolation device, especially for a variety of low-frequency vibration. Besides, the effective isolation frequency is fixed which cannot adapt to the variation of the operation condition excitation source. Therefore, it is of great practical significance to propose a method and scheme which can effectively restrain the low-frequency vibration of pipeline system of NPP. In order to meet the above requirements, an electromagnetic semi-active vibration absorber is designed in this paper. Its electromagnetic stiffness can be adjusted by changing the excitation current, so as to achieve the purpose of changing the natural frequency online and improving the vibration attenuation effect. In order to verify the rationality of its electromagnetic structure, the magnetic field distribution of vibration absorber is simulated and analyzed. Results show that it is close to magnetic saturation inside without leakage. On this basis, a platform of simulated condensate water pipeline system for vibration experiment is set up, on which the frequency response together with time & frequency-domain vibration characteristics of typical support points are tested, and the stiffness, damping and other parameters of the vibration absorber are optimized according to the test results. Finally, the vibration of pipeline system is simulated and compared. The results show that under the condition of the optimal combination of parameters, the vibration spectrum of the pipeline is obviously reduced after the vibration absorber is used, which indicates a significant vibration control effect.

The high-temperature gas-cooled reactor pebble-module (HTR-PM), which is under constructed in Shidao, Shandong Province of China, adopts the spherical graphite coated components as fuel elements. During the operation of the reactor, the spherical fuel elements will be reloaded into the core periodically through the refueling pipelines, which will cause radioactive irradiation to personnel and facilities in several corridors and rooms. The three-dimensional geometry model was built according to the real structure and the radiation shielding design was verified and optimized by the QAD-CGA program. In the calculation, the average gamma intensities of the spherical fuel elements of 15 different cycle times were adopted as the source term. From the results, it can be concluded that: 1) For the 20-meter corridor, the dose rate will be less than 1E−6mSv/h at 30cm outside the concrete wall of 0.6m thickness by single fuel elements; 2) For the 0-meter corridor, an extra 32.0 cm steel shield should be added to satisfy the personnel dose rate limit of 0.03mSv/h; 3) For semiconductor electrical equipment closed to the pipelines, the required steel shield thickness is 12.0 cm under the accumulative dose limit of 50Gy; 4) For non-radiation resistant cables, the required steel shield thickness is 4.5 cm under the accumulative dose limit of 1000Gy.

The paper presents a Relap5 study of the influence of the centrifugal pump characteristics on the dynamic loads on piping system after power failure. Interpolated and experimental pump characteristics are used. The differences between the interpolated and measured pump curves, the general description of Relap5 model and results of calculations in form of selected time curves for rotational speed, volume flow, pressures and dynamic forces are presented and discussed. The analysis of the results shows that the maximal dynamic force on pipe section calculated with experimental pump curves can be up to 6 % higher than respective calculated using interpolated curves. However, it is not possible to determine to what extent the differences are caused by the interpolation itself or caused by the differences in the design of the centrifugal pumps. The latter since it differs more than 50 years between the pumps whose characteristics are used for interpolation and the pumps with corresponding experimental characteristics.

The orifice throttle pipelines with large pressure drop suffer damages from two aspects: high frequency vibration caused by cavitation and low frequency vibration caused by high flow velocity. In order to solve the vibration phenomenon of typical throttling pipe with large pressure drop in the system of nuclear power plant, the key hydraulic characteristics such as pressure drop, flow velocity, streamline and eddy current were simulated and analyzed for single-stage orifice throttling pipe. The negative pressure area was found in the downstream of the orifice, where cavitation occurred. Eddy current was formed due to the large local velocity caused by the orifice plate jet. Then, the throttle performance of multi-stage concentric orifice plates was evaluated by means of blockage pressure drop method. The cavitation damage was relieved greatly. But it was not eliminated, especially in the last stage orifice. The expanding type five-stage orifice plate was designed according to pressure drop stage-decreasing principle, with which the possibility of cavitation was eliminated but large pressure drop resulted in large flow velocity at the downstream of the first orifice plate. Multi-stage eccentric orifice plate was designed with the consideration both in cavitation characteristics and velocity distribution, which could eliminate the harm of cavitation and reduce the low frequency vibration caused by large flow velocity to a maximum extent. As a result, multistage eccentric orifice plates could be recommended as an optimized design scheme for the vibration control of the orifice piping with large pressure drop.

Pressurizer surge lines are essential pipeline structure in NPPs, and the thermal stratification in surge line is recognized as one of the possible cause of thermal fatigue. In this paper, a Computational Fluid Dynamic (CFD) method has been adopted to simulate temperature fluctuations on the process of temperature rising in a pressurizer surge line under rolling motion of single degree of freedom. This work focuses on a fundamental description of differences of thermal stratification between the surge line rolling around the coordinate X-axis condition and that in a static state. The Large-eddy simulation (LES) model is employed to capture the details of temperature change in surge line. Temperature distributions near the inner wall of a surge line pipe with or without swinging were monitored and compared. The temperature differences between the top and bottom of the pipe sections are employed to represent the maximum temperature differences at all the monitored sections. As the surge line swinging, the pattern of temperature distribution and the length of thermal stratification development are different from that in a static. Fluid temperature fluctuation in surge line occur periodically during the fluid temperature rising when the surge line is rotated with the X-axis, and the temperature difference between top and bottom of the surge line is reduced in the same motion mode compared with the static state.

Water hammer phenomenon is a significant pressure wave in pipe system caused by momentum change when the moving fluid is forced to stop or change direction instantaneously. Common causes of water hammer are sudden valve closing at the end of a pipeline system, pump failure, check valve slam etc. The steam transportation pipeline system may also be vulnerable to water hammer when it confronts with the situation where liquid and steam co-exist. Water hammer often occurs when steam condenses into water in a horizontal section of steam piping. Then steam “picks up” water to form a high-velocity “slug” and create extra stress to pipe. When steam is trapped into sub-cooled water, the collapse of vapor cavity can lead to collision of two columns of liquid, resulting in a large rise in pressure which will damage pipes, supporting structures and hydraulic machinery. Nuclear power plant is composed of complex equipments and piping systems, lots of which contain both liquid and steam. Hence, there is a potential threat of occurrence of water hammer to the normal operation of systems. Thus, this phenomenon needs to be well investigated and prevented with some effective methods. For the purpose of overpressure relief under severe accidents, the spent fuel pool cooling system of CAP1000 series nuclear power plant provides a discharge passage from containment to spent fuel pool. When the containment pressure exceeds the control value, valve is opened to discharge high-temperature and high-pressure steam until the pressure drops to a safety value. During this process, serious water hammer happens, causing pressure rise beyond the design pressure and further leading to damages to pipes and structures. Therefore, water hammer of overpressure discharge pipeline in CAP1000 plant is studied in this work. On the basis of verification of the capabilities of computational code RELAP5/MOD3.3, hydraulic transient of water hammer is simulated under different conditions. It is indicated that after steam discharge stops, residual steam in pipe condenses because of contact with sub-cooled water in spent fuel pool. Subsequently, the rapid backflow and vapor cavity lead to a severe water hammer. The detailed analysis has shown that water temperature of spent fuel pool has a decisive influence on the mechanism of water hammer phenomenon, including collision of liquid column to valve disc and cavity collapse in the horizontal pipe. The collision and separation of liquid column result in relatively lower pressure amplitude.

Under the conditions of the actual operation of the nuclear power plant, especially in the drainage system of the heaters, water recycling system, and condensate pump recirculation system, there is a sudden pressure drop after the control valve in the pipeline, which will lead to serious vapor-liquid two-phase flow. Vapor liquid two phase flow can lead to accelerated corrosion in the pipeline. In order to study the influence of vapor liquid two phase flow on the accelerated corrosion of the pipeline, an experimental facility for simulating the operating conditions of the Second Loop was designed and completed. The experimental facility comprises a pressure regulator, high pressure water tank, circulating pump, power heater, low pressure water tank, condensing and cooling system, instrumentation and data acquisition system. The flow characteristics and the corrosion characteristics of the inner wall of the pipe in the flow channel after the control value were studied in the experiments. The flow characteristics include the distribution of water and vapor temperature, pressure drop distribution and void fraction along the flow channel. In the meanwhile, the corrosion rate and surface characteristics of inner wall at different locations were also studied by CCD and SEM. In addition, the influences of pipe material and pipeline geometry on the erosion-corrosion were also studied in the experiments.

The present paper provide an available FEA method for the structural analysis and seismic design of pipelines coated with flexible shielding materials, in nuclear power plant, subjected to earthquake action. The modal and spectral analysis are carried out by using FEM software ANSYS. By comparing the numerical simulation results of dynamical behavior of the pipelines with or without flexible shielding materials, the effects of the flexible shielding material on the seismic performance of the pipeline system in nuclear power plant are discussed. Finally, a local supporting constraint for reducing seismic effects is put forward, which can significantly decrease the maximum stresses in the pipelines coated with flexible shielding materials.

In the pipeline system of nuclear power plant (NPP), the transfer of pump induced vibration is quite noticeable, which will not only affect the operation stability of the equipment, but also cause the noise pollution problem. In this paper, an active vibration control (AVC) device is designed to reduce the pipeline system vibration induced by condensate water pump operation. This device consists of controller, sensor, actuator and conditioning circuit, and an adaptive control strategy is adopted. The pump rotating speed detected by the tachometer is transferred to the controller as the reference input signal, which is used to excite the actuator to generate an opposite force acting on pipelines and attenuating the vibration passing to foundation structures. To avoid the influence on pipeline system compactness, the structure design of actuator is adaptive to pipe clamp. Then a prototype is manufactured and a verification experiment about the vibration control effect is carried out on the stimulant bench of feed water system. At the design speed of condensate pump, the pipe vibration characteristics at various flux conditions are compared and analyzed. The results indicate that the vibration response of pipe clamp is well restrained when the AVC device is installed. Especially, the outstanding characteristic line spectrums of pump source are significantly declined; therefore, the total vibration level of the pipe clamp is effectively controlled.

The security of pipe system is an important problem in nuclear power plant. The displacement field and vibration frequency of the pipeline system under given loads and displacement boundary conditions are calculated with isogeometric analysis. Comparisons with commercial softwares are given to prove its effectiveness. In isogeometric analysis the non-uniform rational B-spline (NURBS) basis functions are used to accurately describe the geometry of pipeline, which can be imported directly from the geometry model designed by computer aided design (CAD) tools for numerical analysis. The quarter bend pipes are described precisely with only limited control points and the approximation by straight line elements in finite element method based simulation is replaced. Present simulation with isogeometric analysis can simplify the modelling process of pipeline system and improve the efficiency of the simulation with great accuracy.

For nuclear power stations, the main function of Essential Service Water System (ESWS) is to discharge the waste heat from reactor core and spent fuel pool to the environment controllably, which is directly related to the safety and economy of nuclear power stations. Usually ESWS use open water from sea, rivers, lakes, reservoirs, as heat transfer medium. Extremely harsh environmental conditions may disable system functions and even lead to ESWS failure, directly reduce the safety and economy of nuclear power stations, and cause serious nuclear accidents. Failure of ESWS is one of the main reasons that lead to the Fukushima nuclear accident because of the loss of electricity after the earthquake and tsunami. Based on the typical ESWS configuration and conditions of serving nuclear power stations in China, the influence of environmental conditions on the function of water system is studied, and the corresponding measures are analyzed. These conditions can be divided into three categories: temperatures, water levels, and physical and chemical characteristics. Temperatures affect cooling characteristic of ESWS mainly. Nuclear power stations in tropical areas need to focus on cooling capacity might be reduced by high temperature. Those in cold region need attention to excessive cooling and freezing problems caused by low temperature. The influence of water levels is mainly fluid transport capacity and selection of equipment to ESWS. When the range of natural water level is too wide, designers shall consider measures to narrow it, such as the construction of highly reliable reservoir. Inland nuclear power stations shall try to ensure the reliability of ESWS; prevent water level changes beyond the scope of design caused by drought and flood disasters. The effects of physical and chemical properties are derived from the open water characteristics, including high salinity, high chloride ion concentration, carrying solid particles, suspended solids, and aquatics, and so on. These characteristics will cause the equipment and pipeline eroded or even damaged, aqueducts of intake and output jammed, heat exchangers of the final heat sink weakened and other negative effects, resulting in ESWS performance decline. Some of these factors are the characteristics of station site natural environment, some others are changes caused by human activities. Some factors are sustained, long-term; some others may be sudden, temporary. Influence on these factors need to be taken measures from many aspects, including structure, biological disinfection, special materials and equipment, environmental protection measures around the nuclear power station, and so on. On the whole, the environmental factors that affect ESWS in the nuclear power stations are wide, and the influence mechanism is more complex. These factors ultimately act on ESWS, but most of them cannot be banished inside of ESWS or the final heat sink system. Against the negative effects from environmental conditions, it has to be considered from all steps in the engineering of nuclear power stations, including design, construction and operation. All the measures shall be suitable to local conditions, in order to ensure the safety and economy of nuclear power stations.

For nuclear power plant life extension projects, safety related buried steel pipelines need to be assessed for the effects of postulated accidental impact loads. Examples of such pipelines include emergency firewater lines and control room vent ducts. Published solutions concerning the buried pipeline beneath a surface impact, often aim at a generic problem dealing with a pipe within a half space. The interface between the buried pipes and the underground structure is often ignored. In the case that a buried steel pipe penetrates into nearby underground structure in short distance from a surface impact, shear and bending may originate from reactions at pipe-structure interface, in addition to ovaling and buckling of the pipe section. Neglecting such boundary effect may lead to underestimating potential damage of buried pipeline, and jeopardizing the safety of the nuclear power plant. Comprehensive analysis of such structure-pipe-soil system is often subjected to availability of state-of-art finite element tools, as well as costly and time consuming. This study presents a simplified approach using basic concepts, in order to evaluate the buried pipeline beneath a heavy surface impact considering nearby pipe-structure interface. Applied impulse loadings for postulated surface impact are first established, followed by the evaluation of consequential vertical pressure directly on the top of the pipeline. The pipe cross-section is checked for ovaling and buckling. The closed form solution is then proposed to evaluate effect of pipe-structure interface.

Pipelines are widely used in many fields including power industry, petroleum system etc. Pipelines such as the surge line and main pipe are easily subjected to thermal stratification as a result of the non-uniform temperature distribution in the nuclear power plants. Furthermore, pipelines can suffer from thermal fatigue in virtue of long-term uneven stress distribution. When the surge line or main pipe subjected to thermal stratification and thermal fatigue keeps operating for long time, the pipe leakage may happen because of the existence of pipeline crack. The thermal pipeline crack leakage mainly appears in the region with stress concentration. As the pipe system is always covered with thermal insulation layer in the actual nuclear power plants, it is hard for workers to observe pipeline leak, which can have a bad effect on the normal operation. Since the temperature and humidity close to the pipe crack due to leakage can change compared to the normal operation, we can infer from the temperature and humidity changes that the pipe leakage occurs. Based on this idea, the temperature and humidity near the crack of the pipe need to be measured to detect the leakage fields. As the fluids with high pressure and high temperature flow in the pipe system in an actual nuclear power plant, the pipe leakage experiment was performed in the high pressure and high temperature condition. When the fluids with high temperature and pressure leak in the crack, the water will evaporate quickly, which means this process belongs to spray flash evaporation process. The temperature and humidity variations were monitored in the experiment with temperature and humidity probes which have the advantage of responding to the change of temperature and humidity sensitively. The data collection program was mainly written based on the LABVIEW platform. The collecting time step was set 1s. As the measuring position and leakage flux are two key factors for the pipe leakage, the experiment was carried out with different measuring positions and leakage fluxes conditions. The experimental results showed that the leak flux had an important influence on the temperature and humidity near the pipe crack. The temperature and humidity started to change in a very short time with large leak flux. At the same time, the velocity of the temperature and humidity change was high with large leak flux. When the pipe leakage occurred in the location near the temperature and humidity probe, the temperature and humidity responded quickly and the velocity of temperature and humidity change was large. The experiment data can be used for the prediction of the pipe leakage in the nuclear power plants.

The Chemical and Volume Control System (CVCS) is one of the most important systems in nuclear power plants to ensure the pressure stability of the primary loop. CVCS consists of various equipments, valves and pipelines; the operation of the system is relatively complex. The sneak circuit analysis technology was applied to identify the patent path and temporal sequence in the CVCS system design and running. Improvements and advices are provided to lower the potential risk.

Thermal stratification can occur when hot and cold fluids mix inside the pipeline of nuclear power plant, which can lead to temperature fluctuations with large amplitude and high frequency with time-varying wall heat stress and even induce thermal fatigue phenomenon. Therefore, how to accurately get temperature fluctuations of the inner wall without damaging the whole structure of the pipe system becomes the prime problem to be solved in the study of thermal fatigue. Base on energy conservation law, the non-iterative mathematic model of inverse heat conduction using the volume-control method has been developed in this paper for pipe in nuclear power plant. Numerical and experiment results show that the volume-control method can accurately catch the temperature fluctuation of different positions on the inner wall.

Pipe rupture in the Condensate Extraction system (CEX) of nuclear power plants will lead to high oxygen content in the secondary circuit and therefore exacerbates equipment and pipeline corrosion. At the same time, there is a risk of loss of vacuum in the system, which has a direct impact on the safe and stable operation of nuclear power plants and will affect the economic benefits. In this paper, Equipment Failure Root Cause Analysis (ERCA) methodology is employed combined with metallography analysis (SEM analysis, XRD) and finite element simulation analysis, to investigate the root cause for drainage pipeline rupture in CEX System of Pressurized Water Reactor (PWR). Detailed analysis process of ERCA was introduced including RCA project establishment, data collection, failure modes analysis and so on. The most probable failure mode is pointed out through the investigation and evidence analyzed. It suggests that the improper design and the installation of limiting orifice plate should be the root cause. And corresponding corrective actions are put forward in details to prevent the recurrence.

Fatigue is the major aging mechanism for pipelines and components that are subject to transient during plant operation. In order to avoid abrupt accidents caused by fatigue failure, it is necessary to adopt corresponding techniques to monitor components condition due to fatigue damage. Experience shows that on-line fatigue monitoring system is the most effective way to monitor fatigue damage status of pressure-retaining boundary subjected to operational loads in real-time for NPPs. Several commercial fatigue monitoring systems have been put into application in the nuclear power plants. In this paper, the difference of the international fatigue monitoring systems were compared. As the key technologies of fatigue monitoring system development, the processing of operational load data and the stress influencing function development at monitored locations were researched in the paper. Furthermore, in order to study on the methodologies of the key issues of the system development, the experimental facility was established to improve the knowledge on fatigue monitoring system development for two kinds of input data modules, and the reasonability of the different methods in the study was proved reciprocally. Finally, the on-line fatigue monitoring system for pipelines was established to evaluate actual fatigue damage status.

Nuclear power plants (NPPs) typically have four classes of electrical supply systems to provide redundant and resilient power to ensure safe operation, cooling, and shutdown of NPPs. Class III emergency power traditionally uses standby diesel generators to fulfill equipment electrical demands. This requires a reliable source of diesel energy and reliance on one particular kind of fuel source for backup power. During the North American blackout of 2003, hospitals had to run on backup generators that were mostly diesel fired. A prolonged blackout could have negatively affected hospital infrastructure and patient health. With current diesel generators, capable of supporting critical systems for up to 72 hours (for most NPPs), a prolonged loss of power or transient could similarly also have adverse effects on NPPs. A Combined Heat and Power (CHP) system is a power generation system that generates electricity, in addition to concurrently having heating and cooling capabilities. Often, gas or steam turbines are used to generate electricity. CHP heating and cooling capabilities can be met via absorption coolers and heat pumps. Low-grade exhaust heat from turbines could be used for the absorption processes. Absorption coolers/heaters produce cold/warm fluid using a heat source via the vapor compression cycle, taking advantage of high affinity fluids. The absorbent allows for the refrigerant to boil at lower operating conditions, which allows for heat transfer. CHPs’ that are capable of accepting natural gas as a form of fuel will increase resiliency of NPP power supply. This ensures a reduced risk of running out of fuel during prolonged transients due to continual supply via pipelines already in place. The objective of this work is to generate a conceptual design of a multi-fuel source CHP system that is capable of at least accepting natural gas as an alternative to Class III diesel generators for a NPP. The system will be capable of supplying Class III power to Small Modular Reactor (SMRs), as well as commercial NPPs, such as CANDU, PWR, and BWR types. Current state of the art Class III backup power systems, CHP systems, CHP thermodynamic cycles, multiple compatible fuels, as well as absorber heaters and coolers have been investigated. Ranking systems were used to determine the top three designs. The parameters included turbine type, efficiency, flow rates, operating temperature/pressure, fuel type, fuel to energy ratio, fuel availability, absorber chiller/heater fluid efficiency, operating temperature and thermal conductivity. Based on the ranking system parameters, a thermodynamic model including mass, energy and entropy balance of an 8 MW e CHP with heating and cooling capability between the ranges of 15°C to 27°C was conceptually designed.

As a safety device to alleviate the loss of reactor coolant, the siphon breaking system is widely used in nuclear power plant. Researchers are very interested in this technique for its “passive” characteristic. Vertical downward air-water two-phase flow is encountered in the siphon breaking process. Previous researches have been more focused on some physical parameters, such as water flow rate, air flow rate, pressure drop and the undershooting height. Void fraction, as a key parameter in multiphase flow, should be studied in the siphon breaking phenomenon. Therefore, a needle-contact capacitance probe is used for flow-phase identification and a single-wire capacitance for obtaining the average value of gas distribution along the straight line. Experimental results show that the flow pattern during the vertical downward air-water two-phase flow is mostly annular flow. With the gas entering the pipeline, void fraction profile against time can be divided into three stages. The slope in the first stage is similar to that in the third. However, the slope slows down in the middle stage. The experimental results also show that the real duration time to break the siphon flow is as short as about 6 s. The void fraction at the end of the siphon breaking process is about 0.38. During this stage, a large amount of gas is sucked into the downcomer and little water is inhaled. The gas phase results in a convergent effect, where the air intake is the direct and fundamental reason of siphon breaking.