bluuebunny/crossref_metadata_2025 huggingface dataset:
The Hot Test of Primary Separation Equipment of Steam Generator for CAP1400 Nuclear Power Plant
Authors: Bingbin Ying , Xilin Li , Feixi Zhang , Shaoxuan Lin
Published: December 2017

A new type of Steam separator for CAP1400 Steam Generator is introduced. The result of characteristic experiment, the variables of which contain steam load, saturated water flow and water level, shows its separation characteristics are suitable for CAP1400 Steam Generator. Steam generator is a major equipment of pressurized water reactor power plant, one function of which is to transfer heat from reactor to feedwater and to generate qualified steam. Steam separator, located in upper shell of Steam generator, is used to separate steam from saturated water/steam mixture which exits from the tube bundle region.

Steam separating process usually consists of three stages. Centrifugal separation first happens in Primary separator, where most of water is separated. Gravity separation is achieved in the space formed between Primary separator and the dryer. Finally, as the relatively low moisture content mixture flows through the chevron-type dryer, the qualified steam is obtained.

The new developed primary separator, named SP3 type, consist of raise cylinder, swirl impeller, downcomer and vane. As the steam/water mixture enters the primary separator, a centrifugal force is induced by the swirl. Thus the greater dense water in the steam-water mixture is forced toward the outside diameter of the separator, while the less dense steam stays in the central portion of the separator. The separated water leaved the separator through the downcomer. Relatively high quality steam exits the separator, passing upward through the center of raise cylinder.

The paper introduced the SP3 primary separator developed by hot test, which includes test device, device design, test method, and result analysis. Separation characteristics of SP3 primary separator are obtained by various tests. The result of varying steam load test shows the outlet moisture is 13.3% under the condition of 100% steam load and 4.3 recirculation ration. The result of varying saturated water flow test shows the outlet moisture increases as the saturated mixture flow rate increases. The outlet moisture increased slowly as saturated water flow rate increases when steam load is at 50% and 100%, while the outlet moisture increases fast when steam load reaches at 120%. The result of varying water level test shows the water level is not a major factor for the SP3 primary separator performance.

By analyzing experimental data, a fitting formula of the primary separator performance is obtained. the steam flow rate is the most sensitive parameter affecting primary separator performance in that formula. The water flow is the subordinate parameter for the primary separator performance. the resistance characteristics of separator is also obtained, which supports the research and development of the CAP1400 steam generator..
Experimental Study on Active Vibration Control for Feed Water Pipeline System of NPP
Authors: Wei Zhiguo , Dai Lu , Wang Junrong , Li Shaodan , Xiao Qi
Published: December 2017

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.
Investigation on DDC Susceptibility of 690 Nickel-Alloy Welding Wire
Authors: Gu Yu , Zuo Bo , Zhang Junbao , Huang Yifeng , Yu Yan
Published: December 2017

690 nickel-based welding materials are widely applied on nuclear equipment manufacturing. While the nickel-based welding metal bath has the properties of high viscosity and poor fluidity, resulting in high susceptibility to welding cracking of the nickel-based weld metal, especially ductility dip cracking. As the nuclear key equipments serve in a severe condition combined with high pressure & temperature and corrosive atmosphere, the existence of micro crack would greatly affect the life expansion of the equipments. In this paper, WHS690M -alloy welding wire was developed in order to meet the requirement of CAP1400 nuclear power plant, which has good welding performance. Strain-to-fracture (STF) test based on Gleeble-3500 thermo-mechanical simulator was performed to evaluate the DDC susceptibility of Inconel 52M and WHS690M weld by different welding parameters. And combined with EBSD, the characteristics of grain orientation was also analyzed. The minimum critical strain in STF test appeared at 1050°C, and the value reached 3.2% for WHS690M while 2.1% for Inconel 52M. Furthermore, the critical strain for both WHS690M and Inconel 52M groove weld metal was about 2.0%, which was lower than cladding metal.
Extended Study of Reactor Coolant Pump Coastdown Phenomenon With RELAP/MOD3
Authors: Zhong Y. , Xia D. , Zhou W. X.
Published: December 2017

Mutual effects of loop and pump characteristics should be considered for reactor coolant pump (RCP) inertia design. Based on pump four-quadrant characteristics, numerical model of RCP coastdown analysis was developed to evaluate the combined effect of loop and pump characteristics. Flow coastdown was influenced by pump factors such as inertia and friction loss, and loop factors such as friction loss and form loss. However, speed coastdown was dominantly relevant to pump factors rather than loop factors. Comprehensive effect of all the factors was defined as Initial Kinetic Energy Ratio, ε. Flow coastdown relative delay to speed coastdown was linear related to ε. This delay shall be fully considered for those loops with higher ε, to avoid excessive margin for pump inertia design which relates to lower overall efficiency and significantly increased design difficulties.
Research on the Shock Resistance Optimal Design of Submerged Nuclear Power Plants Model Subjected to Underwater Explosions
Authors: Zhang Wenzheng , Feng Zhipeng , Lv Xi , Xiong Furui , Song Haiyang
Published: December 2017

The research on the linear and nonlinear responses of the plant structures, especially with internal equipment, subjected to underwater shock load make great significance in assessing the shock resistance ability of submerged nuclear power plants. In this paper, an optimal design of a small-sized submerged vehicle with internal components has been carried out to enhance the shock resistance. The dynamic responses and deformation mode of the vehicle subjected to underwater shock wave are firstly analyzed to determine the weak parts. The incident wave induced by UNDEX is equivalently replaced by half-sine shock wave, as the calculation of fluid-structure interaction (FSI) in the optimization process is very complicated. Based on the constraint condition of invariant weight, an optimization procedure is established by means of the commercial finite element code and optimization method. Numerical simulation is carried out to validate the effective of the optimal procedure. Compared with the original one, the maximum Mises stress of the submerged vehicle decreased from 426MPa to 359MPa, approximately reducing by 15.73%. The optimal result can be used to redesign the submerged vehicle with internal components suffered from underwater shock loading, as well as to enhance the anti-shock capability of submerged nuclear power plants.
Large Scale PWR Passive Containment Cooling System Wind Tunnel Test
Authors: Xinxin Pan , Jingyu Huang , Chunjing Song
Published: December 2017

CAP1400 is a large passive pressurized water reactor nuclear power plant, which relies on engineering safety features such as passive containment cooling system (PCS) to remove the decay heat in the containment and mitigate accident consequences. PCS is designed to perform passive containment cooling which is mainly dependent on natural convection inside the containment and inner wall condensation heat transfer, outer containment surface water film coverage and evaporation heat transfer and external air flow path cooling performance, etc. Among them, the key factors that affect the performance of the external air flow path include the flow resistance characteristics of the air flow path and the wind-direction neutrality characteristics. The relevant performance will be the important design input of the accident analysis, which will directly affect the safety of the power plant.

During the normal operation of power plant, the PCS air flow path is influenced by the external environment, and its internal flow is very complicated. Designers are often lack of data support, and can’t fully consider the impact of environmental flow. In order to fully study the performance of PCS air flow path, it is necessary to perform PCS integrated scaled wind tunnel test.

According to the original design of CAP1400 PCS system, the model scale research is developed and CAP1400 PCS wind tunnel test scaled model is established and the scale is 1:100. The test model includes shield building model and the surrounding plant model, which contain pressure measuring points uniformly distributed in 6 horizontal cross sections of the shield building. The pressure measuring point arrangement does not affect air flow in the air flow path. The following wind tunnel tests are simulated in different wind speed including 15m/s, 20m/s, 10m/s, 25m/s. The air flow pressure, wind velocity at the inlet and outlet of air flow path and the pressure distribution of inner annulus and outer annulus are measured in order to study the air flow pressure drop and wind-direction neutrality characteristics, and the wind tunnel test also considers the different wind direction angle, with and without the surrounding buildings and the effects of different landforms.

The test results show that the flow rate of inlet and outlet of air flow path is balanced and the wind velocity at the upwind and central area of the flow path outlet is larger than other area, and a large vortex comes on the leeward side near the wall. The local uneven flow phenomenon exists in the outer annulus of the air flow path, but the wind pressure distribution of inner annulus is not affected by environment wind speed, wind direction angle, landforms and the surrounding buildings. So CAP1400 PCS air flow path has the characteristics of wind direction neutrality, and the natural convection of the air flow path will not be adversely affected by the environment wind.
Seismic Performance of PCS Water Storage Tank Considering Fluid-Structure Interaction
Authors: Gou Hongliang , Shang ZiDuan , Sun Yugang , Chu Meng , Ge Honghui
Published: December 2017

Water sloshing of PCS water storage tank (PCSWST) can cause a significant effect on the dynamic response of Shield Building under seismic loads. It is complicated to perform the calculation of water sloshing especially for the tanks with irregular shapes. Consequently, it is important to establish an appropriate equivalent mechanical model for simulation[1], [2].

In this paper, the water sloshing is firstly investigated based on the potential flow theory, which including the seismic modal analysis. Based on the theoretical research, a highly efficient (simplified) calculation formula is derived, which mainly considering the impulse mass, convective mass, position function and spring stiffness etc., through this way the equivalent model for PCSWST is established by applying mass-spring element. The equivalent models based on Housner & Graham theory[3], [4] are also established. Additionally, the 3-D finite element model of water sloshing considering fluid-structure interaction is established by using the software of Ansys.

Total of four models are built as shown in the paper, then modal analysis and dynamic response under earthquake excitation are performed using ANSYS. The results are compared to justify the equivalent model in this paper. The results indicate that Graham formula did not provide the correct location expressions for the convective masses. The expressions for the impulsive mass and its position given by Housner are not satisfactory. As a comparison, the results from the equivalent model, which is recommended in this paper, can best fit the data from finite model. From above results and comparisons, a more reasonable and refined equivalent model for PCSWST design is provided.

Based on the equivalent model established, the influence on structure caused by the increase of water mass is analyzed. The results from the seismic analysis are compared, including member force, shear strain and shear force. Based on the research, the feasibility of the design is analyzed, which can provide important support for the structural design.

Finally, the seismic reduction of water tank is studied using the finite element model established in this paper. The horizontal and vertical anti-sloshing baffles are designed. The maximum acceleration and displacement corresponding to different baffle length are compared to study the effect of the seismic reduction.
Study on the Design Methodology of Swirl Nozzle for Nuclear Reactor Pressurizer
Authors: Cao Xiong , Ding Zhiwei
Published: December 2017

Pressurizer is one of the most important components in reactor coolant system of a nuclear power plant, which operates normally at pressure of 15.4 MPa and temperature of 345°C[1]. The main function of pressurizer is to regulate the pressure in the reactor coolant system by either cooling the steam or heating the saturated water in its upper zone. When the pressure in the reactor coolant system increases, it will distribute cold water to decrease its temperature and pressure through atomizing the reactor coolant with swirl spray nozzle in pressurizer. Swirl nozzle is the key part of pressurizer with swirl structure of full cone spray pattern, and the atomization performance include drop size, spray angle and distribution, also it is characterized by huge flow rate and low pressure drop, and its atomization performance decides the quality of pressure control of the reactor coolant system. To enhance the independent design level of both pressurizer and cooling system, it’s necessary to study the atomization performance of swirl nozzle for nuclear reactor pressurizer.

Aimed at improving atomization performance of swirl spray nozzle, the structure design methodology of nuclear reactor pressurizer was studied systematically in three aspects including theory design, numerical simulation and test confirm in this thesis. Through designing the swirl nozzle structure according to similar design formula of spray nozzle in theory, especially studying the influence of different structures that mainly include internal swirl structure on internal flow field of swirl nozzles, the primary structure parameters of swirl nozzle were confirmed. Then, through numerical simulation of the internal flow field, flow rate and pressure drop, and swirl core structure of the swirl nozzle (by building physical model and mathematic model according to the spray nozzle structure), the atomization performance of the nozzle was analyzed. On this basis, the typical swirl nozzle was designed and tested, which included spray angle, flow rate as well as pressure drop tests, and spray drop tests, and the applicability of the computational fluid dynamics (CFD) method was verified when it was applied in swirl nozzle design. Finally, the design method of swirl nozzle with deep groove of swirl core for pressurizer was put forward.

Through this studying of theoretical calculation, numerical simulating and test, the correlation between the structural parameters of swirl nozzle and atomization performance was achieved, meanwhile design, analysis and test methods of spray nozzle with low pressure drop and huge flow rate were established. It is helpful to realize the independent design of pressurizer’s swirl nozzle and even to put forward the design methodology of pressurizer’s swirl nozzle with our own characteristic.
Technology of Protection Against the Dynamic Effect of Nuclear Pipe Rupture
Authors: He Feng , Yuan Feng , Ai Honglei , Wang Xinjun , Lu Xifeng , Lv Yongbo
Published: December 2017

The special safety facilities and important equipment, etc. of the nuclear power plant will be damaged due to the whipping nuclear high-energy piping after the rupture, and more serious further damage will be caused. In this paper, the process and method of protection analysis of the nuclear high-energy piping rupture have been given from four aspects. The four aspects are location of high-energy piping break, the jet thrust, whipping behavior analysis, and protection analysis of whipping. On the basis of the traditional energy balance method, the method is improved by considering the energy absorbed by the plastic hinge of the piping and the change in the direction of the jet thrust. And then, the comparisons among the traditional energy balance method, the improved energy balance method, and the 3-D finite element dynamic method have been carried out. The deformation of the whip limiter analyzed by the traditional energy balance method is 20.31% larger than which analyzed by the improved energy balance method, and the deformation of the whip limiter analyzed by the 3-D finite element dynamic method is 30.59% smaller than which analyzed by the improved energy balance method. For the first time, a 3-D finite element model according to the true arrangement of the pipe and the whip limiter model are built to simulate the process of whipping not in the plane, considering the energy dissipation of the whip limiter. For the pipe whipping not in the plane and protecting against the pipe rupture by whip limiter, there is no good way to carry out the protection analysis of the piping rupture in the past. Now, the problem can be solved by the 3-D finite element dynamic method.
Design of the Area Radiation Monitoring System in Nuclear Island of HTR-PM
Authors: Xie Feng , Kong Jiaji , Liu Hongqi , Cao Jianzhu
Published: December 2017

The area radiation monitoring system can supply information of the abnormal radiation levels in a nuclear power plant quickly, give an alarm on site explicitly, and display the dose rate value and alarm information remotely. Thus, it can help the radiation protection officer to take proper actions timely to protect personnel from excess irradiation. According to the experience and technology from the pressurized water reactors (PWRs), the area radiation monitoring system in nuclear island of high temperature gas-cooled reactor pebbled-bed module (HTR-PM) which is a modular high temperature gas-cooled reactor demonstration power plant in China has been designed. It is used to monitor the external radiation exposure dose rate at the passageway in the control areas, operational areas, equipment rooms, etc. The design scheme of the area radiation monitoring system in nuclear island of HTR-PM is presented, which can be divided into two parts, the stationary area γ radiation monitoring subsystem and portable radiation dose rate monitoring subsystem. The operational principle of the system, the layout of the area radiation monitors, the main characteristics of the instruments, and the communication logical framework of the system are introduced in this paper.
Sealing Structure Design for Normal Residual Heat Removal Exchanger
Authors: Li Xilin , Jiao Ming
Published: December 2017

Normal residual heat removal exchanger (RNS HX) is used to remove reactor residual heat during the second stage of reactor shutdown. A hypothetical over-pressurized accident, during which RNS HX would suffer the normal operating pressure and operating temperature of RCS should be considered during the design of RNS HX.

This paper introduces Bolt-Flange-Gasket seal structures for RNS HX, considering the leakage-proof during over pressurized situation and some other operating-related aspects. At tube side, Metal-to-metal contact (MMC), with which a mass energy could be stored in metals during bolt preloading and could release when inner pressure increases, is used, so that the energy stored in metals could resist high pressure to prevent the joint disjoin. Another feather of tube side joint is that a slight angle is designed at the periphery of tubesheet.

At shell side, a segmental gasket, which is divided into two or more pieces rather than a whole circle, is used to seal the joint. As the gasket pieces are installed from the side of tube bundle, the shell does not need to be removed away completely from the bundle in order to install a circle gasket, which could facilitate gasket changing process and save the operating space.
Research and Development of Monitoring and Control Platform in Active Magnetic Bearing
Authors: Xunshi Yan , Yan Zhou , Jingjing Zhao , Zhe Sun , Zhengang Shi
Published: December 2017

In order to cut cost and simplify redevelopment, the paper proposes a new design for monitoring and control system in active magnetic bearing. Data sampling and processing module is separated with data display module. Singlechip processor is used to sample data from the DSP controller, and processes the data into the needed form. A touch-screen only for display takes on the processed data to customers. The design makes the systems easy to develop and shrink the cost to one tenth of previous design.
Research on the Structure Integrity Analysis of Flywheel of Reactor Coolant Pump in Nuclear Power Plant
Authors: Zhang Jian , Zuo Yongde , Chen Xingjiang , Zhang Qinghong
Published: December 2017

The flywheels on reactor coolant pump motors provide inertia to ensure a slow decrease in coolant flow in order to prevent fuel damage as a result of a loss of power to the pump motors. During operation at normal speed, a flywheel has sufficient kinetic energy to produce high-energy missiles and excessive vibration of the reactor coolant pump assembly if the flywheel failed. Overspeed of the pump rotor assembly during a transient increases both the potential for failure and the kinetic energy of the flywheel. The safety consequences could be significant because of possible damage to the reactor coolant system, the containment, or other equipment or systems important to safety.

Usually, the design of connection between flywheel and pump rotor has two types, one is keyway, and the other is rotor shrink fitting. This paper has done the research on the analysis of the integrity of flywheel in design rules and guidelines, such as NUREG, RG and NB, which have given the allowable stress limits but not given the potential for failures types and the stress verification type. So the stress verification of the flywheel is different in different technicals and structures. Some papers also have different analysis methods in China.

This paper considers the failures mode of flywheel, using the analytical method in THEORY OF ELASTICITY and Tresca criteria to give the method of the stress analysis of flywheel. Next, this paper pays attention on the analysis of the flywheel integrity about two connection types, and gives the other requirements of integrity. Such as stress analysis, fatigue analysis, ductile failure analysis, non-ductile failure analysis, crack propagation analysis etc.
Piping Vibration and Vibration Damage Prevention Through Screening of Dynamic Susceptibility
Authors: Zeng Lingfu , Jansson Lennart G. , Börjesson Alexander
Published: December 2017

This paper continues our recent discussion on piping vibration and practical measures for preventing vibration-related damage in nuclear power industries. Our emphasis is on an empirical approach, which attempts to estimate a so-called “dynamic susceptibility” at various locations in a piping system. This approach uses a dynamic susceptibility factor (DS), which is a quotient of the modal stress to the modal velocity, as an indicator of the risk levels of vibration to measure the vibration sensitivity for excitation sources in a given frequency interval of particular interest. In the present paper, Benchmark examples tested by the general purpose finite element program ANSYS, and commercial piping programs CAEPIPE and PIPESTRESS, are presented and the potential of using DS parameters as a screening tool for determining “potentially-large” alternating stresses is illustrated. It is demonstrated that, combined with knowledge of typical vibration sources, this is a practical and cost effective way for forming a base for the vibration control prior to installation and for the planning of post-installation vibration monitoring of a piping system under operation.
Research and Application on Visual Aided Cable Design System
Authors: Ying Zhang , Xiao Yu
Published: December 2017

Nuclear power plants have many types of equipment, dense structure, large plant area, and very complex cable channels. Considering the division and type of cables, cable design has been one of the most complicated tasks in power plant design. Shanghai Nuclear Engineering Research and Design Institute (SNERDI) uses Intergraph PDS system for power plant design, for the AP1000 and CAP series of projects, including equipment and raceway modeling, and use Shaw Cable Manager (SCM) system for cable laying. Cable design work involves multi-system data processing, though without a data integration platform based on 3D visualization. In this paper, a visual aided cable design system is built to display the design data and cable data in a unified 3D plant model, which realizes the visualization and integration of multi-system information, and improves reliability and efficiency of the design work. This method can be applied to a variety of design systems, and has good scalability.
Analysis and Optimization of Heat Treatment Process for Lower Core Support Plate and Core Barrel of Reactor Vessel Internals
Authors: Liu Dongan , Lin Shaoxuan , Ding Zonghua
Published: December 2017

Lower Core Support Plate (LCSP) and Core Barrel (CB) are key components of reactor vessel internals. Especially, since the fuel assemblies are installed on the LCSP, its flatness is critical for the safe operation of fuel assemblies. However, for SM1 and HY1 nuclear power plant (NPP), after heat treatment of the weld between LCSP and CB, the LCSP deforms seriously and its flatness exceeds the limitation, which results in a time-consuming and costly reprocessing. A numerical model of heat treatment process between LCSP and CB was developed first. The general rules of temperature and deformation distribution of LCSP and CB were obtained. Also, an experiment was conducted to validate the model. With the validated model, the deformation mechanism of LCSP due to heat treatment is studied. At last, the heat treatment process between LCSP and CB was optimized to avoid similar issues for the following NPPs.
The Demonstrated Function of PLR Pump Mechanical Seal Under SBO Conditions
Authors: Yoshioka Toshiyuki , Motani Takashi , Fujimaru Atsushi , Yamada Koji , Hasegawa Kunihiro , Shimizu Takeshi , Torigoe Hiroyuki
Published: December 2017

Generic Safety Issue 23 (GSI-23) by the U.S. Nuclear Regulatory Commission (NRC) was identified in 1980 as a result of staff concerns about reactor coolant pump (RCP) seal failure during Station Blackout (SBO), that is, seal degradation leading to a significant loss of reactor coolant in pressurized-water reactor (PWR) plants. Resolutions of GSI-23 have been considered at PWR plants. In 2000, NRC decided to close GSI-23 and issue Regulatory Issue Summary00–002 (RIS 00–002)[1], based on considerations such as; for example, the improvement of RCP seal performance[2], and the reduction of the risk of RCP seal failure in certain plants by the addition of alternate power sources. After the closure of GSI-23, some licensees were planning to make other associated improvements under their individual plant program. In Japan, the RCP seal was showed that leakage rate was low under SBO testing conditions[3] in licensing safety reviews conducted according to new nuclear regulatory standards after the nuclear accident at the Fukushima Daiichi Nuclear Power Plant.

On the other hand, a boiling-water reactor (BWR) is not included in GSI-23 because operating experience indicates that seal failures in BWRs result in smaller leak rates than seal failures in PWRs and, BWRs have the reactor coolant injection capability under SBO conditions, such as the reactor core isolation cooling (RCIC) and the high-pressure coolant injection (HPCI) system. In addition, for the particular BWR-2 type plants that do not have emergency makeup systems, the pump mechanical seal was tested under SBO conditions and successfully showed minimal leakage[4]. However, for the BWR-5 type plants which have the reactor coolant injection capability, such as the RCIC and HPCI systems, the pump seal had not been tested. In Japan, after the nuclear accident at the Fukushima Daiichi Nuclear Power Plant, licensing safety reviews of BWRs and PWRs are being conducted according to new nuclear regulatory standards. We took this opportunity to test the leak rate from Primary Loop Recirculation (PLR) pump mechanical seals under SBO condition.

The peak of leak rate was approximately 0.6ton/h (2.6gpm) during the 24 hours of SBO testing condition. Despite damage of O-rings in the mechanical seal by heated water which were ovserbed at post-inspection test, a very low leak rate was realized because the leakage path after passing through the damaged parts of the O-rings was limited by the other restricting pathway in the mechanical seal. This seal leakage was very low, compared with the reactor coolant makeup capability of the RCIC system and the reactor coolant release capability from main steam safety relief valve (SRV). Therefore, we reconfirmed that the result of this leak rate does not affect the safety evaluation for the reactor. It is shown in this paper that the leak rate from PLR pump mechanical seals is low under SBO condition by our demonstration test.
Research of the Rod Drop Time Based on the Control Rod System of TMSR-SF1
Authors: Lin Zuokang , Zhu Lin , Zhao Chunfeng , Cao Yun , Wang Xiao
Published: December 2017

The reactor scram function realized by the rapidly dropping of control rods ensures safety when the reactor accidents (loss of electricity and earthquake, etc.) happen. In the thorium base molten salt reactor (TMSR - SF1), the rod drop time is obviously affected by the resistance which produced in the molten salt as its high density and viscosity. In this paper, the drop time of the control rod is obtained by the theoretical and experimental methods for comparison. Firstly, the drop time is analyzed both in air and water condition with calculation and experiment. And the method used for the resistance calculation of the rod during dropping is verified. Secondly, the similarity criterion is adopted to calculate the drop time in molten salt condition. The study shows that: 1) In air and water condition, the calculation is coincidence with the experimental results within the maximum error less than 2 %. 2) The drop time of the rod in molten salt is 2.8 s with a dropping height 2.4m in reactor, which satisfy the safety requirement of the control system. 3) It is necessary to use another buffer beside the disc spring to protect the driving mechanism of the rod during the rod dropping.
Numerical Research of CAP1700 IHP Cooling by Natural Convection
Authors: Zhang Xingliang , Yu Hao , Zhang Wei , He Yinbiao
Published: December 2017

The feasibility of using natural convection, in substitution of the cooling fans inside the integrated head package (IHP), to cool the control rod drive mechanism (CRDM) assemblies of the proposed CAP1700 design is studied in this paper.

The flow and the temperature fields of the CRDMs and their surrounding areas, including the IHP, all cooled by natural air convection, were simulated using CFD. The results showed that the average temperature of the CRDM wall surface will stay equal to or below 300°C, which is the pre-requisite of using the newly developed high temperature coils. Therefore, cooling by the natural convection is judged to be feasible for the CAP1700 no-fan design.
Study on Spent Fuel Pool and Storage Racks of the Advanced Nuclear Power Plant
Authors: Yun Ling , Chao Zhang , Lei Li , Hao Qian , Honghui Ge , Zufeng Xia
Published: December 2017

The present paper is concerned with the problem of modeling dynamic test. The study focuses on the fluid-structure interaction (FSI) in spent fuel pool, which is filled with liquid and spent fuel components. Through the comparison of the results of numerical analysis (Based on LS-DYNA) and modeling test, such as the sloshing wave amplitude; the displacement & inclination angle of the storage racks; pressure exerted on the plate of racks and the walls of the pool were obtained, This paper evaluates the seismic performance and the safety margin of the structure, and also provides the surface pressure input for the design of storage racks.

The phenomena of liquid sloshing in spent fuel pool presents a great interest for civil and nuclear engineering, the dynamic effects of sloshing in spent fuel pool could affect the plates of storage racks and the wall of spent fuel pool during strong earthquakes. Study of sloshing in spent fuel pool is proved challenging due to the presence of strong flow interactions with storage racks in the pool. The movement effect caused by the sloshing load of liquid is the primary reason for the problem to become more complicated.

In this study, two models were created, one is scaled model for shacking table test, another is a numerical model utilize finite element analysis code LS-DYNA. LS-DYNA command *CONSTRAINED_LAGRANGE_IN_SOLID was used to simulate the interaction between solid material and fluid material (fluid-solid coupling effect). Considering the fluid-solid coupling effect between solid and fluid, ALE (Arbitrary Lagrange Euler) element and Lagrange element were used to simulate fluid material and solid material respectively based on their different characters in the numerical model. Using the three spatial components of artificial time histories generated from the design response spectra. The following parameters should be paid attention to in both the scaled model shacking table test and the numerical simulation of time history analysis: The relative displacement between the structure of the spent fuel pool and storage racks; inclination angle of the storage racks; pressure exerted on the plate of storage racks and the wall of pool and the sloshing wave amplitude.

In this paper, the integrated pressure represented by the impulsive hydrodynamic pressure on the plate of storage racks between the shacking table test model and the numerical model were calculated and compared. The analytical parameters of numerical model were corrected to make the numerical model result consistent with modeling test result. In this way, a more reasonable numerical model is obtained.
Corrosion Susceptibility in High Temperature Liquid/Vapor Environments of Materials for Tubing of Heavy-Water Collection System in CANDU Nuclear Power Plant
Authors: Li Guangfu , Zhao Liang , Yang Xinghong
Published: December 2017

Corrosion especially pitting on the inner surface was regarded as the first step of the failure process of 316L and 304L stainless steel tubes of heavy-water collection systems in CANDU reactors of a power plant. In this work, four materials including the 316L, 304L, carbon steel A106B and Ni-based Alloy 690 were tested in 14 designed liquid/vapor LiOH-containing environments at both 250 and 150°C, to obtain basic information on the corrosion susceptibilities as a function of temperature, media state, Cl− and Li+ contents, for materials selection. Results showed that the corrosion susceptibility rank were A106B at the top, 690 at the bottom and the stainless steels between them. The corrosion susceptibility was lower in a liquid solution than in the vapor above the liquid, and was relatively lower at 150°C than at 250°C for the same media. Chloride promoted corrosion significantly but LiOH showed some protect effect. In the Cl-free solutions at 150°C, A106B exhibited almost no corrosion in the liquid but localized corrosion in the vapor above. It corroded apparently when adding 3%NaCl into the solution or raising temperature to 250°C. The stainless steels showed no corrosion in both the liquid and vapor states of Cl-free solutions at both 150 and 250°C, but exhibiting pitting in the vapor when 3%NaCl added to the solution. 690 always exhibited excellent corrosion resistance during long term tests in various environments.
The Fracture Behavior of Hydrogen-Charged A508III Low Alloy Steel in the Process of Natural Hydrogen-Escaping
Authors: Xiao Qian , Lu Zhanpeng , Chen Junjie , Peng Hao
Published: December 2017

The effects of hydrogen charging as well as natural hydrogen escaping on fracture behavior of A508III steel at room temperature were investigated. The presence of hydrogen caused a slightly increase in yield strength and a distinct decrease in the elongation and reduction in area, but no or hardly any influence on the ultimate tensile strength (UTS). Non-charged specimen exhibited a ductile mode of failure with micro-void coalescence, while the H-charged specimens showed typical brittle fracture with “fish eye” appearance. The presence of hydrogen gave an increase in the dislocation density after tensile deformation. Nucleation of defects occurred primarily from the inclusions which contain complex oxides of Al, Mg and Ca, and (Ca, Mn)S, as well as MnS only to a small extent. The parameters of non-metallic inclusions, such as type, shape, size and content play the decisive role at the defect formation. In the process of natural H-escaping, the fracture surface characteristics of H-charged steels gradually transferred from brittle fracture mode to ductile fracture mode. At the same H-charged time, the H-escaping time required for the brittle-ductile transition was related to the H-charged current density.
Data Compiling and Analysis of Oxide Film Growth on Nickel-Base Alloys in High Temperature Water
Authors: Ru Xiangkun , Lu Zhanpeng , Chen Junjie , Peng Hao , Xiong Qi , Ma Jiarong
Published: December 2017

A logarithmic relationship for the oxide film thickness of Nickel-based alloys in high temperature water versus immersion time or dissolved hydrogen (DH) is identified based on the compilation and analysis of the experimental data reported in some literature. A theoretical equation is used to quantify the correlations between thickness and time or DH. The logarithmic correlation between inner layer thickness and iron content in alloy matrix and the linear correlation between outer layer thickness and iron content in alloy matrix are primary identified by fitting the experimental data in our own work. The correlation of oxidation kinetics and stress corrosion cracking growth rate is discussed.
Stress Corrosion Cracking Behavior in Liquid/Vapors of Boiling Magnesium Chloride Solution of 304L/316L/690 for Heavy-Water Collection Tubing in CANDU Nuclear Power Plant
Authors: Li Guangfu , Zhao Liang , Yang Xinghong
Published: December 2017

Stress corrosion cracking (SCC) in high temperature water/vapor environments with chloride concentrated as impurity has been regarded as one of the main failure mechanisms for leakage events occurred in 304L and 316L stainless steel tubing of heavy-water collection system in CANDU reactors according to failure analysis. In this work, the SCC behaviors in high temperature environments with highly concentrated chloride of the 304L, 316L and Ni-based alloy 690 were characterized by performing tests with U-bend specimens in the liquid of boiling magnesium chloride solution, as well as in the vapors above the liquid, to provide basic data for selecting better materials. Effect of surface state was examined by testing the specimens with inner surfaces in as-received state and in grinding-polished state for comparison. Results showed that the stainless steels 304L and 316L were all susceptible to SCC both in the liquid and in the vapor above. No apparent difference in cracking susceptibility was found among the specimens with as-received original surface state and with grinding-polish surface state. All were mainly intergranular cracking. The 304L appeared more resistant to SCC than the 316L. Ni-based alloy 690 exhibited excellent resistance to the cracking, as no any cracks were found on all the specimens tested. However, general corrosion occurred on the specimens of 690 tested both in the liquid and in the vapor above.
Design of In-Containment Refueling Water Storage Tank of Chinese 3rd Generation Nuclear Power Plant
Authors: Yulin Liu , Xiaoying Sun
Published: December 2017

In this paper, the structure configurations of the in-containment refueling water storage tank (IRWST) of Chinese 3rd generation nuclear power plants (NPPs) was described firstly. Then, the general structural calculation for several loads, especially thermal load, were presented, as well as the stability evaluation of IRWST base-slab. The effect from fluid-structure interaction was also considered in the calculation to evaluate the design margin of IRWST. Finally, structure strength evaluation was performed for construction load case.
Research on Rolling-Sliding Integrated Auxiliary Bearing and its Application in High Temperature Reactor
Authors: Jianqiang Chen , Zhe Sun , Guojun Yang , Xingnan Liu , Zhengang Shi
Published: December 2017

The active magnetic bearings (AMB), with the advantages of no friction, no abrasion, no lubrication and active control, is used in the primary helium circulator for high-temperature gas-cooled reactor (HTR). But the magnetic bearing is a complex system, which contains sensor, controller, power amplifier circuit and actuators. Any part of failure is likely to make high-speed rotor off balanced position and fell in the inner ring of the bearing, causing huge impact and fiction heat that may damage the magnetic bearing. Therefore, it is necessary to bring the auxiliary bearing in the magnetic bearing to protect and temporarily support the high-speed rotor. The auxiliary bearings are mainly divided into two categories: rolling bearings and plain bearings. Generally speaking, for rolling bearings, the force of friction is smaller and the heat caused by it is lower during the touchdown. However, it needs to be detected online to ensure that it can work well in emergencies, and the rolling bearings has a smaller load capacity because of the point contact between the ball and the ring. Compared with the rolling bearings, the structure of plain bearings is simple and durable. With a larger load capacity and the advantage of non-contact detection, the plain bearing is gradually becoming a research hot-spot in the field of the auxiliary bearing. But the great friction and inevitable heat are also cannot be ignored.

In High temperature gas-cooled reactor demonstration power station (HTR_PM), the work load of helium main fan is very large, once the support of electromagnetic bearing is out of work, the auxiliary bearing need to suffer from a very large drop impact load, which is accompanied by a huge friction fever. Therefore, it is important to develop a rolling-sliding integrated bearing which can bear heavy load and have little friction, combined with the advantages of plain bearings and rolling bearings. And that is an important direction of the development for the main helium pan in high temperature gas-cooled reactor nuclear power plant.

This paper establish a simulation model for a horizontal rotor and rolling-sliding integrated auxiliary bearing system. In the case of synchronous rotation of the inner ring with the rotor, the speed of the outer race of the bearing is determined. and based on the main helium fan in HTR-PM, using the finite element analysis software LS-Dyna, the rolling-sliding integrated auxiliary bearing is proposed and the impact force and the MISES stress nephogram when the peak inflation occurred during the first impact and the axial axes displacement curve during rotor drop in the auxiliary bearing are preliminarily simulated in this paper., then certain theoretical reference is provided for the design and engineering application of the rolling-sliding integrated auxiliary bearing.
Seism Analysis of Large-Type Radioactive Liquid Tank
Authors: Shoutong Dai
Published: December 2017

There are many large standing liquid tanks for collecting radioactive waste liquid in nuclear engineering, the tanks must keep structural integrity and the structural break shouldn’t happen in order to avoid the radioactive leaking accident, so the accuracy of the seism analysis is most important. The liquid sloshing is very sharp in the large standing liquid tank because the weight of waste liquid in the tank always exceeds ten tons.

In order to think over the sloshing effect accurately in seism analysis the 3D mass-spring finite element model is applied instead of the conventional method such as the fixing mass method because the latter is too conservative for Large Radioactive Liquid Tank.

In this paper the software ANSYS is used to simulate the seism response, the waste liquid in the large tank is divided into two portion, one is the static portion and the other is the sloshing portion comparing to the tank on the condition of the seism, the spectrum method is adopted to research the response of the structure model under the seism.

The results show that the mass-spring model referenced to 3-D model is effective for the seism analysis of the large standing liquid tanks. The results of the mass-spring model are more accurate than that of the conventional method.
Design of AP/CAP Construction Cover Module
Authors: Te Tang
Published: December 2017

The reactor building of AP/CAP nuclear power plant is constructed open top. Open-Top Construction can reduce construction time, but will result the reactor building easily affected by bad weather. The two self-reliance projects, Sanmen and Haiyang, use a temporary cover to help the reactor building to withstand bad weather. However, the effects of using this temporary cover were limited. Based on the project logic, construction method and lessons learned, a special cover support is invented and several temporary cover designs are developed. The new developed cover is named AP/CAP Construction Cover Module.
Preliminary Analysis of Pressure Bearing Capacity for a Typical Prestressed Concrete Containment
Authors: Zhou Xuan , Yuan Kai
Published: December 2017

In this paper, the pressure bearing capacity of containment is analyzed to support the severe accident management guidelines development. The finite element model of containment is established. The concrete, the steel liner, and the prestressed tendons models are developed. The ordinary steel bars are in an integrated model with concrete. The prestressed tendons and concrete are connected by constraint equations. Self-weight case, prestressing load case, and internal pressure loads cases are analyzed. Results show that in the self-weight case, the containment is compressed slightly, which can improve the pressure capacity of the containment. In the prestressing loads case, the containment is much more compressed comparing with the self-weight case. In the internal pressure cases, the high stress areas occurred around the equipment hatch.