核动力工程

Flow Visualization Study of Steam-Water Two-phase Flow Phenomena in a Saturated Porous Structure with Internal Heat Source

Zhang Zhen, Yan Xiao, Xiao Zejun, Wang Xiong, Chen Bingde, Zhou Huihui

2019, 40(2): 1-5. doi: 10.13832/j.jnpe.2019.02.0001

Abstract(613)

Abstract:
An visualization experiment was conducted to study the steam-water two-phase flow phenomena in a saturated porous structure with internal heat source. The porous structure was formed by stacking uniform spheres. The two-phase flow phenomena in porous media were obtained by high-speed camera system and the flow regimes was identified. The influence of pressure and inlet subcooling on the flow regimes was also investigated. The results of visualization experiment show that the flow regimes in porous channel are bubbly flow, bubbly-slug mixed flow, slug flow, slug-annular mixed flow and annular flow. The results also show that the bubbly to transitional flow transition and the transitional to annual flow transition appear at lower superficial steam velocity in the channel with the lower inlet subcooling. The present study also show that the bubbly to transitional flow transition and the transitional to annual flow transition appear at lower superficial steam velocity in the channel with the higher pressure.

Model Verification of Al₂O₃ Nanofluid Critical Heat Flux Mechanism

He Xiaoqiang, Yu Hongxing, Wang Jinyu, Jiang Guangming, Dang Gaojian

2019, 40(2): 6-9. doi: 10.13832/j.jnpe.2019.02.0006

Abstract(593)

Abstract:
To overcome the shortcomings of the present models, considering the contact angle and capillary wicking effects, an Al₂O₃ nanofluid critical heat flux(CHF) mechanism model has been developed. In this study, the developed CHF mechanism model is verified by several Al₂O₃ nanofluid and de-ionized water experiments. The verification shows that the present model can simulate the Al₂O₃ nanofluid experiments well, and overcome the shortcoming of the Kandlikar’s model. Also, this model can predict the trend of CHF versus nanofluid concentration, which is a new function that other models do not have. This model also can simulate the de-ionized water experiments well, and the calculated results are similar to the results calculated by El-Genk and Guo model based on de-ionized water CHF experiments, which means that the developed model has a wide applicability.

Simulation of Reflux in U-Tube of Steam Generator Based on CFD Method

Sui Zengguang, Yang Jun, Yang Ye, Dong Shichang, Xu Lejin

2019, 40(2): 10-15. doi: 10.13832/j.jnpe.2019.02.0010

Abstract(592)

Abstract:
Under the condition of natural circulation in reactor systems, reflux may occur at U-tubes of steam generator(SG). Taken a SG of a Generation III Pressurized Water Reactor (PWR) with passive safety systems as the prototype, the thermal-hydraulic characteristics of the SG reflux are investigated. In this paper, the SG U-tubes are divided into six groups according to their lengths, and the flow and heat transfer characteristics of single-phase fluid in the U-tubes are analyzed with (CFD) methodology. The trends of total pressure drop against mass flow rate are derived based on the simulation results. The effect of the U-tube length and the temperature difference between the primary side fluid and the secondary side wall(?T) on the reflux are discussed. It is shown that when ?T is constant, the reflux occurs more often in the longer U-tubes. When the U-tubes length is constant, the smaller ?T is, the more often reflux occurs.

Effect of Pebble Bed Structure on PB-FHR Neutron Physics Characteristics

Ji Ruimin, Yan Rui, Chen Xingwei, Yu Shihe, Zhou Bo, Zou Yang

2019, 40(2): 16-22. doi: 10.13832/j.jnpe.2019.02.0016

Abstract(199)

Abstract:
In the Pebble Bed Fluoride-salt-cooled High-temperature Reactor(PB-FHR), the pebbles are loaded by buoyancy in the coolant. Hence, the packing structures of the pebbles in the core are greatly influenced by various factors such as the core structure, the loading rate, the coolant density, and the coolant flow. The uncertainty of the pebble bed structures is the key issue in the reactor design and safety analysis. To provide a mock up for the Solid Fueled Thorium Molten Salt Reactor (TMSR-SF1), the Pebble Recirculation Experiment Device(PRED) was developed. Packing experiments has been performed under different conditions. For the purpose of verifying the design, the study on pebble structures has been performed based on the TMSR-SF1 design using Monte-Carlo code MCNP. The results quantitatively evaluate the effect of various packing fractions, bottom shapes of the pebble bed and the loss of coolant on the key neutron physics parameters. It shows that, when the packing fraction is increased from 50% to 64%, the effective multiplication factor, the pebbles loaded and the temperature coefficients are increased, but the control rod worth is decreased; the bottom shape of the pebble bed has different effects depending on the structures. The positive cone can cause both increasing and decreasing of the reactivity, and the inclined plane and the negative cone cause a decreasing of the reactivity; the loss of coolant causes a denser packing which results in a decreasing of the reactivity.

Hydraulic Analysis for HPR1000 Reactor with PHYCA Program

Tang Huapeng, Liu Yu, Chen Xi, Huang Huijian, Shen Caifen, Li Song

2019, 40(2): 23-26. doi: 10.13832/j.jnpe.2019.02.0023

Abstract(544)

Abstract:
For the requirements of the reactor hydraulic analysis, the PHYCA program has been developed by Nuclear Power Institute of China on self-reliance. In order to evaluate its engineering adaptability, the analysis of the reactor hydraulics of HPR1000 was performed based on PHYCA. Pressure drop in the reactor vessel, bypass flow and hydraulic load of the reactor internals were calculated. The results of PHYCA have good agreement with the engineering design values. It demonstrates that PHYCA is suitable for the analysis of the reactor hydraulics of similar nuclear power plants.

Study on Post-Irradiation Microstructure of UO₂ Pellets in Recycled Uranium Fuel Elements

Peng Yanhua, Zhu Wei, Wen Bang, Yang Fan, Fan Shen, Meng Zhiliang, Jiang Linzhi

2019, 40(2): 27-31. doi: 10.13832/j.jnpe.2019.02.0027

Abstract(314)

Abstract:
In order to evaluate the irradiation stability of UO₂ pellets in the recycled uranium fuel elements, the axial and radial distribution of irradiation swelling, crack, grain size distribution and grain growth behavior of the irradiated UO₂ pellets were observed and analyzed by means of a hot cell metallographic microscope. The results show that a large amount of cracks occur in the fuel pellet and obvious circular distribution present in the cracks of recycled uranium, and radioactive divergence distribution present in the natural uranium. For both the recycled uranium and natural uranium, the grain size of fuel pellets is equiaxed, grain size grows gradually from the edge region to the center region of pellets, and grain boundary become coarser after irradiation. No obvious difference in grain size, morphology and distribution between the recycled uranium and natural uranium has been found. Additionally, the irradiation swelling of the recycled uranium pellets is not obvious, and no significant difference in the crushing degree of the pellets and grain growth process between the recycled uranium and natural uranium has been found under the same operating conditions in the reactor.

Inhibition Effects of Low Concentration of Boron on Corrosion of Zirconium Alloy

Zhao Yongfu, Tang Min, Jiang E, Wen Juhua, Yin Zhaohui, Liu Jinhua, He Yanchun

2019, 40(2): 32-36. doi: 10.13832/j.jnpe.2019.02.0032

Abstract(264)

Abstract:
The high temperature electrochemical corrosion behavior of zirconium alloy was studied by potential dynamic polarization and electrochemical impedance spectrogram (EIS) under three normal boron-lithium water chemistry conditions with the same pH and the uniform corrosion behavior of zirconium alloy was investigated by weight gain and microanalysis under two lithium-concentrated water chemistry conditions. The high temperature electrochemical corrosion tests indicate that the increase of boron concentration can reduce the passive current density and increase the electrochemical impedance value of zirconium alloy under three normal boron-lithium water chemistry conditions despite of the same pH, consequently the boron is beneficial to corrosion inhibition of zirconium alloy. The uniform corrosion tests show that adding boron can obviously decrease the corrosion weight gain, oxide film thickness and compact the oxide films by contrast to the lithium-concentrated water chemistry condition without boron, thus the boron is beneficial to improve the corrosion resistance of zirconium alloy under lithium concentrated water chemistry conditions.

Creep Analysis of Reactor Pressure Vessel Lower Head under Core Meltdown Severe Accident

Luo Juan, Luo Jiacheng, Li Pengzhou, Sun Lei, Tang Peng

2019, 40(2): 37-41. doi: 10.13832/j.jnpe.2019.02.0037

Abstract(713)

Abstract:
During the core meltdown severe accident in the nuclear power plants, the strategy of In-Vessel Retention (IVR) to contain the melt in the Reactor Pressure Vessel (RPV) is a key mitigation measure. During the implementation of IVR strategy, the RPV lower head is likely to fail due to excessive creep deformation under the combined action of extremely high temperature loads and mechanical loads. Therefore, it is necessary to perform the analysis of the creep deformation of the RPV lower head, to ensure the structural integrity of the RPV under the condition of melt retention. In this paper, under the assumption of IVR, the finite element method is used to perform the thermal-structural coupling analysis of the RPV lower head. The temperature and stress fields of the vessel wall, and the plasticity and creep deformation of the lower head are calculated. Combining the plasticity and creep rupture criteria, the failure was analyzed. Results show that the deformation of the structure will be greatly increased when creep is considered. During the IVR strategy under severe accident, the main failure mode of the RPV lower head is creep failure instead of plastic failure. Further analysis implies that the internal pressure has a significant influence on the creep deformation and failure time. This paper provides the method for the creep and failure analysis of the RPV lower head under severe accident.

An Optimization Strategy for High Efficiency and Low Axial Load Design of Hydraulic Model for Nuclear Reactor Coolant Pump

Lu Yeming, Wang Xiaofang, Zhou Fangming

2019, 40(2): 42-48. doi: 10.13832/j.jnpe.2018.06.0042

Abstract(293)

Abstract:
Aiming at the single consideration of the performance characteristics while the lack of the load characteristics in most studies about the reactor coolant pump (the main pump), a new impeller meridian design approach along with the radial basis function neural network and the multi-optimization algorithm was herein adopted, as a result, a new optimization strategy whose targets were the higher performance and the lower axial load was established. To verify the effectiveness of the proposed optimization strategy, it was then applied into the practical design process whose design target was the previous scaled main pump. It could be found as follows: with only three control variables and fifteen samples, the new strategy succeeded in optimizing the pump; in relative to the target pump, the efficiency of the optimal structure is increased by 0.9%, the head is increased by 0.6 m, while the axial load is decreased by about 200 N; the predicting results from the approximate model could quantitatively prove that the increasing of the head can easily lead to the increasing of the axial load.

Research on Active Floating Raft of Pumps

Chen Jiu, Cai Longqi, Liu Jia, Liu Lizhi, Huang Wei, Li Yi

2019, 40(2): 49-52. doi: 10.13832/j.jnpe.2019.02.0049

Abstract(212)

Abstract:
With the increasing requirement of the control of the pump vibration, the vibration control technologies were deeply studied for various pumps worldwide. Vibration control measures such as floating raft and active vibration isolation have been applied in varying degrees. Based on the requirement of the control of the pump vibration, the technology combining the active and passive vibration reduction (the technology for the vibration isolation by active floating raft) is proposed, which is with the centralized arrangement of pumps and the combination of vibration isolation by floating raft and active vibration isolation. It can isolate the broadband vibration and inhibit the characteristic spectrum.. The feasibility of active floating raft vibration isolation was verified by the analysis of natural frequency and vibration reduction effect. The results show that the active floating raft is with good performance in vibration reduction.

Sensitivity Analysis of Effect of Anti-Vibration Bar Gap on Natural Frequency of U Tube in Steam Generators

Tan Wei, Jiang Songyuan, Xiong Guangming, Jia Zhanbin, Guo Kai

2019, 40(2): 53-57. doi: 10.13832/j.jnpe.2019.02.0053

Abstract(228)

Abstract:
In order to meet the requirements of installation and manufacturing, there is a certain gap between the anti-vibration bar and the tube in the part of the U tube of the steam generator. This gap leads to the uncertainty of the restraining strength of the anti-vibration bar, so it is necessary to carry out the sensitivity analysis of the gap. Through the modal analysis in numerical simulation, the spring stiffness is used to characterize the effect of the anti-vibration bar gap on the restraining strength. The results show that when the anti-vibration bar is completely failed, it has great effect on the first natural frequency , which can reduce it by 88.15%. The single point constraint failure has little effect on the natural frequency, and it can only reduce the first natural frequency by 2.65% to the maximum. In each supporting position, the continuous two point constraint near the straight tube has the greatest effect, and the first natural frequency can be reduced by 23.58%.

Fluidelastic Instability of Tube Bundles under Two-Phase Flow Subject to Unsteady Fluid Force

Jiang Tianze, Li Pengzhou, Ma Jianzhong

2019, 40(2): 58-61. doi: 10.13832/j.jnpe.2019.02.0058

Abstract(231)

Abstract:
Fluidelastic instability causes the large amplitude of the heat transfer tube, which results in its wear out. It is the key mechanism of the fluid-induced vibration of the team generator tube bundles subject to two-phase flow. In order to predict more accurately the critical flow of instability, the motion equation of single tube is firstly established utilizing the unsteady fluid force coefficients of two-phase flow which obtained by fitting experimental result data. After the nondimensionalize and Galerkin discretization of the analytical model, the critical flow velocity of each void fraction is calculated by solving the system of equations. Numerical results show that the numerical critical fluid force of instability agrees well with experimental results, which proving that the analytical model utilizing fitting parameters of unsteady fluid force of two-phase flow is available for the predicting of the critical velocity of instability.

Calculation of Natural Frequency and Added Mass Coefficient of Spatial Bend Tube Bundle

Liu Liyan, Xu Wei, Tan Wei, Li Zhao, Guo Kai

2019, 40(2): 62-67. doi: 10.13832/j.jnpe.2019.02.0062

Abstract(302)

Abstract:
It is a crucial step for the safe and stable operation of the intermediate heat exchanger to determine the reasonable values of natural frequencies and added mass coefficients of the spatial bend tubes during the engineering design process. This work aims to calculate the natural frequencies and added mass coefficients of the concentric spatial bend tubes of the intermediate heat exchanger in air and liquid sodium by finite element method. The concentric tube bundle was divided into three typical areas. For these typical areas, three-dimensional models of heat exchanger tubes and fluid fields were built to calculate the natural frequencies and added mass coefficients at six places. The results show that the natural frequency of the heat transfer tubes in the air is less affected by the bending radius of the spatial bend tube, while the added mass coefficient is greatly affected by the bending radius of the spatial bend tube and the place in which it is located. The tubes located in the approximate triangular arrangement place have the lowest natural frequency and the highest added mass coefficient.

Diversity Assessment Method and Its Application in I & C System of Offshore Small Modular Reactor

Yang Yuqi, Wang Yuan, Xiong Guohua, Wang Hongtao, Guo Yongfei

2019, 40(2): 68-73. doi: 10.13832/j.jnpe.2019.02.0068

Abstract(244)

Abstract:
The diversity of I&C system in the offshore small modular reactor is assessed by the standard method of diversity analysis. Aiming at the selection of the platform for the digital I&C system, the diversity quantitative evaluation is conducted by using the method of NUREG/CR 7007, and which provides data support for the selection of the platform for the digital I&C system. The analysis results indicate that the I&C systems of the small modular reactor basically meet the diversity requirements in correlative standards, and are provided with sufficient diversity.

Linear Active Disturbance Rejection Control for Fast Reactor Power and Core Coolant Outlet Temperature

Liu Fengming, Zhou Shiliang, Shen Cong, Wang Xiaoluo, Zhang Huaxia, Liu Yuyan

2019, 40(2): 74-79. doi: 10.13832/j.jnpe.2019.02.0074

Abstract(538)

Abstract:
For the requirements of fast reactor controller with faster response speed and higher control accuracy, Linear Active Disturbance Rejection Controllers (LADRC) are designed for the reactor power and the core coolant outlet temperature, respectively. Based on the fast reactor neutron dynamics kinetics and the core heat transfer equations, the second order nonlinear models of the relative power and coolant outlet temperature for the controller design are derived, and the corresponding Linear Extended State Observers (LESO) with model information are designed. The range of LESO bandwidth is chosen via the time scale parameters of the derived second order model. The range of PD bandwidth is estimated by the allowable range of deviation and actuator operating speed, and the LADRC parameter tuning is carried out accordingly. The simulation results show that the LESO with model information has better total disturbance estimation capacity. The designed LADRC all are with faster control speed and higher precision, and the control performances of LADRC using LESO with model information are better.

Internal Initiating Event Level 2 Probabilistic Safety Assessment during Full Power Operation of Xi'an Pulsed Reactor

Tang Xiuhuan, Wang Baosheng, Zhu Lei, Shen Zhiyuan, Yang Ning, Shan Jianqiang

2019, 40(2): 80-84. doi: 10.13832/j.jnpe.2019.02.0080

Abstract(226)

Abstract:
To quantify and estimate the risk of fission products release from Xi’an Pulsed Reactor(XAPR), the special technical essential of XAPR Level 2 Probabilistic Safety Assessment (PSA) was studied, and accident progression and confinement response were analyzed. Then the source terms analysis of XAPR for internal initial event during 2 MW full power operation condition was performed. The results show that in all release categories (RC), the contribution of confinement intact RC01 which is representative of success sequence is about 97.52%. In all abnormal release categories, the confinement seal failure RC02 gives a high contribution of 75.81%. The results also demonstrate that the highest frequency of all categories is from RC01 which is about 4.80×10^-6/(reactor·year）, but its release quantity is the smallest which is 10⁹~10¹³ Bq. The largest release quantity is about 10¹⁴ Bq from RC06 which is described as confinement seal failure and confinement by-pass sequence, and the release frequency of RC06 is about 1.38×10^-8/(reactor·year）. Based on the results, suggestions are given that the ventilation systems should be shut down to prevent fission products from leaking through the stack and the reactor building.

Priority Management System Scheme Design for ACPR1000 Reactor Protection System Based on FirmSys

Shi Guilian, Zhang Bin, Yang Wenyu, Yin Baojuan

2019, 40(2): 85-89. doi: 10.13832/j.jnpe.2019.02.0085

Abstract(307)

Abstract:
Standard DI&C-ISG-04 and the functional requirements on priority management system (CIS) of ACPR1000 nuclear power unit are reviewed. Based on the experience feedback from CPR1000 nuclear power plant, difficulties in CIS design are analyzed and a scheme of RPS priority management system for ACPR1000 nuclear power plant reactor based on FirmSys is proposed. Compared with that of CPR1000 nuclear power station, the CIS for CPR1000 nuclear power plant is further optimized such as the localized design based on self-reliance, better prevention of software common cause failure, more reliable online periodic test, modular design with multiple interfaces, and onsite function design. and the design sheme has been successfully applied in ACPR1000 nuclear power station.. The practice shows that the CIS scheme proposed in this paper is a feasible scheme.

Study on Monitoring of Action Assembly of Control Rod Drive Mechanism in Nuclear Power Reactors

Xie Ximing, Peng Hang, Zhang Zhuo, Sun Qihang, Zhang Zhifeng, Huo Meng

2019, 40(2): 90-94. doi: 10.13832/j.jnpe.2019.02.0090

Abstract(336)

Abstract:
Because it is difficult to detect the condition of Control Rod Drive Mechanism (CRDM) in reactors by normal detection method, this paper presents one new method to monitor the the condition of CRDM action component by on-line measuring of the coil inductance value. The corresponding relationship between the coil inductance value and the action component has been acquired by experiment, and the condition monitoring has been conducted under the prototype simulation failure condition. The results show that the condition monitoring method can be used for the monitoring of the CRDM action component condition accurately, and the method can be easily implemented without additional equipment requirment in the reactor.

Comparison Study of Boron Detection Method in Boron-Containing Wastewater of Nuclear Power Plant

Jiang Bei, Chen Ding, Li Fuzhi, Niu Tingting, Wang Xin

2019, 40(2): 95-98. doi: 10.13832/j.jnpe.2019.02.0095

Abstract(585)

Abstract:
As an important neutron absorber, a large number of boron exist in the first circuit of nuclear power plant. It is necessary to measure the different boron concentration during the study of the Nuclear power plant radioactive waste. In this paper, four kinds of boron detection methods are compared and their advantages and disadvantages are analyzed. The result shows that national standard of curcumin method is only suitable for detecting low concentrations of boron (<1.2 mg/L); Mannitol titration method is only effective for high concentrations of boron (>1 g/L); Azomethine-H acid colorimetric method is suitable for a wide range of boron concentration and has a high accuracy, but the operation is complicated and the developer need to be kept away from light. The operation of using ICP-OES detecting boron concentration is simple and quick, which is suitable for a wide range of boron concentration. The accuracy and precision can meet the testing and analysis requirements.

Reliability Design and Application of NPP Digital I&C System Based on PSA

Jiang Guojin, Li Fu, Sun Wei, Sun Yongbin, Mo Changyu

2019, 40(2): 99-104. doi: 10.13832/j.jnpe.2019.02.0099

Abstract(686)

Abstract:
Traditionally, the NPP Digital I&C mainly relies on improving the reliability to meet the safety objectives of the plant. With the improvement of regulatory requirements, based on the improvement of equipment reliability, the design method based on probability theory technology has gradually become a new research direction of the safety design for nuclear power plants. Based on the analysis and research of typical power plant initial events, this paper applied the probabilistic safety assessment (PSA) technology to carry out PSA analysis and modeling, and integrate it in the PSA model of the power plant. Through the quantity analysis results, the weak points are identified, and the optimization modification is given. On this basis, a set of reliability design process for realizing the overall safety objective of NPP Digital I&C is proposed to enhance the overall safety of nuclear power plants.

Simulation Study of Mechanical Shim Control Strategy for Large Pressurized Water Reactor

Sun Jian, Wang Pengfei, Yu Yun, Zhang Rui, He Zhengxi, Chen Zhi, Wang Yuanbing

2019, 40(2): 105-111. doi: 10.13832/j.jnpe.2019.02.0105

Abstract(186)

Abstract:
One of the advantages of Mechanical Shim (MSHIM) operation technology is that it can realize partial decoupling between core power and axial offset (AO) in terms of control means. However, the original control strategy design does not make full use of this advantage. A new improved MSHIM control strategy is proposed through theoretical analysis in this paper. At the same time, a simulation platform for MSHIM control system is developed based on the multiple node reactor models. Upon this platform, the original control strategy of Westinghouse, Drudy’s control strategy of Westinghouse and the improved control strategy proposed in this paper are all simulated and compared carefully. The analysis results show that the improved MSHIM control strategy proposed in this paper not only can provide much tighter axial offset control but also can reduce the control rod movement, indicating its superiority than the original operation strategy. The study can be used as a reference for engineering implementation.

Optimization Design of HPR1000 Reactor Lower Plenum Structure

Zhao Wei, Li Yan, Du Sijia, Yu Zhiwei, Xia Xin, Li Hao, Wang Shangwu

2019, 40(2): 112-116. doi: 10.13832/j.jnpe.2019.02.0112

Abstract(637)

Abstract:
In the design of HPR1000 reactor structure, in order to improve the structure safety, the in-core instrumentation detector is guided out of the reactor from the vessel head instead of the lower plenum. The structure of the lower plenum has been changed, which affects the uniformity of the inlet flow field, so a new lower plenum structure is needed. The reactor lower plenum structures of CNP1000, AP1000 and EPR are compared. Based on the characteristics of HPR1000, four schemes for optimum structures are presented. By modeling and CFD analyzing, the four schemes and that for CNP1000 are contrasted and analyzed in terms of structure, manufacture, assembling and flow field analyzing. It is found that the program which uses flow distribution plate is the best, and the uniformity of coolant mixing and flow distribution at the core inlet can be satisfied.

Engineering Analysis and Research on Fit-up for Field Automatic Welding of Primary Piping in Nuclear Power Plants

Dong An, Wang Heng, Guo Lifeng

2019, 40(2): 117-119. doi: 10.13832/j.jnpe.2019.02.0117

Abstract(189)

Abstract:
The importance of the fit-up of reactor coolant pipe and its risk were analyzed. Focused on the fit-up situation of the primary piping, this paper proposes an algorithm for the close fit-up dimension chain, and the initial value, constraint value and optimization value were set according to actual engineering. The predicting of the fit-up situation was realized. Meanwhile, this algorithm was tested and verified by theory and engineering. The result is consistent with the actual value.

Optimum Design of Personnel Airlock Pneumatic Circuit and Improvement of Tightness Detection Procedure for Nuclear Power Plants

Xie Honghu, Zhang Feng, Qin Junwei, Chen Chuyuan

2019, 40(2): 120-122. doi: 10.13832/j.jnpe.2019.02.0120

Abstract(207)

Abstract:
The paper introduces the design of pneumatic circuit for the personnel airlock in nuclear power plants, and presents an optimal design by considering the potential failure modes of the personnel airlock. In terms of the overall accidental tightness failure and the accuracy of the partial tightness detection, the tightness detection procedure of the personnel airlock has been improved. The optimized design of pneumatic circuit and the improved tightness detection procedure have been successfully applied in the third generation nuclear power reactors under construction. It ensures the great enhancement of the operational safety of personnel, and at the same time, the possible tightness failure of the personnel airlock which may result in the loss of integrity of the third safety barrier is avoided.

Analysis and Evaluation for Installation Deviation of Pressurizer Vertical Support Anchor Frames

Deng Feng, Tan Guowei, Li Huanming, Huang Yan, Lu Jia, Hu Yu

2019, 40(2): 123-128. doi: 10.13832/j.jnpe.2019.02.0123

Abstract(235)

Abstract:
Pressurizer vertical support anchor frames in nuclear power plants with pressurized water reactors are primary embedded part. Deviation occurs often during the installing process. It is difficult to evaluate the displacement deviation impact on the follow-up installation of pressurizer, for its complicated structure. This paper analyzed the adjustable margin of the vertical supports of the pressurizer. The optimal installation position of the pressurizer is determined by simulation numerical calculation, based on the actual deviation conditions of pressurizer vertical support anchor frames, and considering the installation requirements of the pressurizer. In order to reduce the difficulty of pressurizer installation on site, this paper put forward the optimization recommendation about the support structure based on the structure characteristics of the pressurizer as follows: increasing the diameter dimension of the lower plate sleeve holes to increase the adjustable margin of lower plate, using the flat washer instead of the spherical washer to relax the perpendicularity requirements of the anchor bolts installation.

Design and Practice of Data Replay in DCS Simulator of Nuclear Power Plants

Ma Jianxin, Peng Li, Dong Xiaofeng

2019, 40(2): 129-132. doi: 10.13832/j.jnpe.2019.02.0129

Abstract(374)

Abstract:
In the training of the nuclear power plant operators, the Distributed control system(DCS) simulator is required to have the data replay function, so that the trainer could monitor the operator's operation. This function is the main difference between the simulator and the actual DCS system, as well as the main difficulty to implement the simulator. This paper fully discussed the design and implementation of the data replay function, and the data replay function is realized through storing and querying the system data, events, mouse and keyboard operation in the system. This technology has been realized in SpeedySim, which is the DCS simulator product for the nuclear power plants, and it has been applied in the full scope simulator system in unit 5 and 6 in Yangjiang Nuclear Power Station. This technology can be promoted for the application of other simulator products and systems.

Analysis of Typical Faults and Treatment Measures of VELAN Pneumatic Valve

Peng Ning, Li Yuexing

2019, 40(2): 133-136. doi: 10.13832/j.jnpe.2019.02.0133

Abstract(253)

Abstract:
The VELAN pneumatic valve experienced typical faults such as valve disc cracked and air leakage from the pneumatic head, during the installation and commissioning of the nuclear power plant project. In-depth analysis and research in this paper show that the root cause of valve disc cracked is the welding quality defect in the valve disc manufacturing process, resulting in increased brittleness, and the valve is subjected to overload stress during use, resulting in brittle cracking, and the treatment is to replace the valve disc assembly and packing, and to conduct a leak test. The root cause of leakage of the pneumatic head is that there is a deviation in the size of the O-ring, and insufficient compression causes leakage, the treatment is to increase the torque of the pneumatic rod center nut, tighten the torque, and then perform the sealing test again after the fastening is completed, or measure the thickness of the compensation ring and the size of the O-ring and process it, such as replacing the compensation ring or O-ring. Through the measures, the typical faults of VELAN pneumatic valves are effectively processed, which provides a reference for the analysis and treatment of similar problems in the future.

Improvement of Ultrasonic Testing Process for Bottom Seal Head and Pipe Plate Circumferential Weld of Steam Generator in EPR Project

Zhao Guobin, Hu Anzhong, Bie Chao, Jiang Shujie

2019, 40(2): 137-140. doi: 10.13832/j.jnpe.2019.02.0137

Abstract(275)

Abstract:
The bottom seal head and the pipe plate circumferential weld of the steam generator in EPR project is the important boundary of the first loop. Because of the large thickness and the spherical shape of the weld, the ultrasonic inspection process should be made with full consideration of the effects such as the attenuation of the probe sound beam and the change of the angle of the probe, so as to meet the requirements of relevant inspection standards. With the in-depth study of the RCC-M and EN 1713 standards, conclusions can be made based on the analysis and simulation of the energy and angle change of the probe, that, by adding a 60 degree probe in the inspection process, the requirements of standard can be satisfied, and the risk of missing dangerous defects can be reduced so that the quality of nuclear safety equipment can be ensured.

Research on 3-Dimensional Dynamic Tolerance Analysis Method for Reactor Control Rod Drive Line

Li Hao, Chen Xungang, Du Hua, Li Yan

2019, 40(2): 141-145. doi: 10.13832/j.jnpe.2019.02.0141

Abstract(276)

Abstract:
One-dimensional linear accumulation method is used in the conventional tolerance analysis for the pressure water reactor driveline, which is too conservative. Limit calculation cannot be used in the conventional analysis method, and there is no guanrantee for 100% confidence level of tolerance design. To solve this problem, a 3-dimensional (3D) dynamic tolerance analysis method for the driveline is proposed. The dynamic change of the drive rod from vertical state to oblique state is analyzed. The first contact point and the second contact point are used as the double reference to restrain the offset direction and range of drive rod and its passage. 3D dynamic tolerance analysis method improves the accuracy of the analysis and reduces the conservativeness of the calculation. The validity of this method is verified by the control rod drive line cold test on the modular samll reactor (ACP100).

Installation Technology for Weirs of Steel Containment Vessel in a Passive Pressure Water Reactor

Zhao Xu, Wang Hongjin, Ding Haiming

2019, 40(2): 146-149. doi: 10.13832/j.jnpe.2019.02.0146

Abstract(178)

Abstract:
By analyzing the structure and function of weirs and adopting effective quality control measures, a weirs installation technology is developed. The installation technology is applied to unit 1 and unit 2 of Haiyang Nuclear Power Plant. The results show that after the installation of weirs system, the design requirements are met and the functional test conditions of the containment cooling system are satisfied. Therefore, the weirs installation technology is effective and feasible.

Research on Sensitivity Analysis of Passive System for Nuclear Power Plants Based on HFR

Zhang Yongfa, Jiang Lizhi, Cai Qi

2019, 40(2): 150-154. doi: 10.13832/j.jnpe.2019.02.0150

Abstract(204)

Abstract:
Sensitivity analysis is applied to uncertainty analysis and reliability evaluation of the thermal-hydraulic process for reactor passive system, which can quantitatively identify uncertainty input parameters that have important influence on system’s thermal-hydraulic behavior. Based on hybrid of random balance designs and Fourier amplitudes sensitivity test (HFR), the passive residual heat removal experimental system of a nuclear power plant is used as an example to carry out the global sensitivity analysis. The applicability and accuracy of HFR method are demonstrated by simulation results. The change rule of parameter importance with time and the parameter importance ranking when the system is running stably are given by sensitivity analysis, which help to guide passive system’s design optimization and operation management.

Analysis of Signal Noise of Supercritical Water Natural Circulation Flow

Ma Dongliang, Zhou Tao Feng Xiang, Huang Yanping

2019, 40(2): 155-160. doi: 10.13832/j.jnpe.2019.02.0155

Abstract(186)

Abstract:
When there is noise in the time series signal of the natural circulation flow, wrong conclusion could be arrived in the calculation and analysis. In order to avoid this result, based on the experimental data signal of the supercritical water natural circulation flow, the denoising of the experimental flow signal is analyzed by selecting various wavelet basis functions. The calculation and comparison of the indicators showed that, the denoised signal value for the natural circulation flow after Dmey wavelet base function transformation are with miminum standard deviation and root-mean-square error (RMSE), maximum correlation coefficient and high noise-signal ratio. The wavelet basis functions is suitable for the analysis and treatment of the signal denoising of the experimental data of supercritical water natural circulation flow.

Application of Functional Reliability Assessment Method in Reliability Analysis of XAPR Reactor Core Natural Circulation

Wang Baosheng, Tang Xiuhuan, Zhu Lei, Bao Lihong, Shan Jianqiang

2019, 40(2): 161-166. doi: 10.13832/j.jnpe.2019.02.0161

Abstract(175)

Abstract:
Functional failure becomes an important factors in natural circulation system operation failure, which needs to be considered in the system reliability analysis. However, the reliability evaluation work of these passive systems is still in the primary stage. In this paper, based on the passive functional reliability evaluation process, the physical process of passive natural circulation was simulated by RELAP5 program and the reliability on capacity of natural circulation cooling in Xi'an Pulsed Reactor (XAPR)reactor core was estimated. Combined with middle loss of coolant accident, the functional failure criteria was established based on the integrity of fuel claddings and the key parameters of natural circulation were identified. According to Latin Hypercube Sampling, sensitivity assessment and functional reliability evaluation were carried out. Then the functional reliability evaluation results can be integrated in the Probabilistic safety accessment (PSA) model. The results show that XAPR PSA not only analyzed the hardware reliability but also considered the functional reliability, to better guide the XAPR operation and improve the safety.

Reactor Core Power Mapping Based on Bayesian Inference

Li Songling, Peng Xingjie, Jiang Zhumin, Yu Yingrui, Li Qing

2019, 40(2): 167-170. doi: 10.13832/j.jnpe.2019.02.0167

Abstract(200)

Abstract:
The reactor core power mapping method based on Bayesian inference has been implemented, and it provides an effective way to combine the information from the measurements of in-core neutron detectors and the numerical neutronics simulation results. Measurements from Unit 1 of Daya Bay Nuclear Power Plant are used to verify the accuracy of the Bayesian inference method, and comparisons are made among the Bayesian inference method, the Kalman filter method and the coupling coefficients method. The root mean square errors, the maximum relative errors, and the power peak reconstruction error of the Bayesian inference method are less than 0.31%, 1.64% and 0.07% for the entire operating cycle, respectively, and the Bayesian inference method outperforms the Kalman filter method and the coupling coefficients method in terms of accuracy. The reconstructed assembly power distribution results and the calculation speed show that the Bayesian inference method is a promising candidate for the on-line core power distribution monitoring system.

Study on Fuel Management Strategy for Irradiation of Small Batch of CF3

Wu Lei, Xu Yang, Lu Di, Li Mancang, Jiao Yongjun, Liao Hongkuan, Yu Yingrui, Li Song

2019, 40(2): 171-175. doi: 10.13832/j.jnpe.2019.02.0171

Abstract(259)

Abstract:
In order to realize the small batch application of third generation China fuel assembly (CF3), the fuel management strategy of the cycle 4 to cycle 7 of the unit 2 of Fangjiashan NPP is studied in this paper. Based on the consideration of the economy, the safety of nuclear power plant operation and the irradiation test requirement of small batch application of CF3 fuel assembly, the fuel management scheme for the small batch irradiation of CF3 fuel assembly is established. In order to further improve the maximum discharging burnup of CF3 fuel assembly, the feasibility of 55000 MW·d/t(U) for CF3 fuel assembly burnup is analyzed. The research shows that the fuel management scheme for the small batch irradiation of CF3 fuel assembly meets the requirement of the safety and economy of nuclear power plant operation, and satisfies the irradiation test requirement of small batch application of CF3 fuel assembly. If the core loading scheme of the third cycle of the fuel management scheme is properly adjusted later, the burnup of 55000 MW·d/t(U) for CF3 fuel assemblies can be achieved.

Dimension Optimization Design of TRISO Fuel Particle in Metal Matrix Microencapsulated Fuels

Xin Yong, Li Yuanming, Tang Changbing, Chen Ping, Zhou Yi, Gao Shixin, Liu Shichao, Zhao Yanli, Yue Huifang

2019, 40(2): 176-179. doi: 10.13832/j.jnpe.2019.02.0176

Abstract(718)

Abstract:
Microencapsulated fuel is the fuel pellet or fuel rod which formed by coated fuel particles dispersed in the matrix, and it is one of the most potential accident tolerant fuel(ATF) fuel. Coated fuel particles can be tristructural isotropic(TRISO) or bistructura isotropic(BISO), and the matrix is metal or ceramics. Based on the ABAQUS software, the metal matrix microencapsulated fuel is calculated. By analyzing the effect of the coating thickness on the fuel performance, suggestions for the optimization are given. The results showed that the thicker the Buffer is, the higher the burnup for the failure is, and thus, the increasing of the Buffer thickness should be considered in the design. The thicker the IPyC layer is, the larger the maximum hoop stress is, and thus, its thickness should be reduced in the design. The thicker the SiC layer is, the smaller the hoop stress is, and thus, its thickness should be reduced in the design. The results of this study could provide a guidance for the optimization design of the metal matrix microencapsulated fuels.

Analysis of Effect of Different Depletion Timestep on k_inf Calculated by KYLIN-II Program

Guo Fengchen, Chai Xiaoming, Lu Wei, Tu Xiaolan, Liu Dong, Chen Dingyong, Zheng Yong

2019, 40(2): 180-183. doi: 10.13832/j.jnpe.2019.02.0180

Abstract(168)

Abstract:
KYLIN-II program uses a projected predictor-corrector method for the critical burnup calculation. In this paper, the effect of different depletion timestep on the calculated k_inf was analyzed for PWRs with pure UO₂ fuel assemble, boron UO₂ fuel assemble and gad-loaded UO₂ fuel assemble, using KYLIN-II program. By comparing with the results, the proper depletion timestep is provided for different types of fuel assembles, to have accurate solution.

Coupled Neutronics and Thermal-Hydraulics Simulation of RIA for Small LBE-Cooled Fast Reactor

Yang Dongmei, Liu Xiaojing, Zhang Tengfei, Cheng Xu

2019, 40(2): 184-188. doi: 10.13832/j.jnpe.2019.02.0184

Abstract(210)

Abstract:
The coupled tool based on neutronics code SKETCH-N and thermal-hydraulics code COBRA-YT has been developed via Parallel Virtual Machine (PVM) software platform. COBRA-YT code performs the thermal-hydraulics calculation and transfers its results such as coolant density and fuel temperature to the neutronics code SKETCH-N to update the cross-section; then SKETCH-N carries out the neutron-physical simulation of the reactor and provides the power density to the thermal-hydraulics code COBRA-YT as boundary conditions. Finally, this coupled code platform is used in the lead-bismuth fast reactor design to simulate some transient and control rod withdrawal accidents. The reactor power increases rapidly and reaches the peak at 1.42s after the control rod withdrawal. Meanwhile, the cladding temperature reaches the maximum 1264℃, exceeding its design limit. The results achieved so far indicates that the control rod withdrawal accident poses a threat to the core with the same enrichment, and the optimization work on the core zoning scheme should be done.

Effects of He Ion Irradiation on Electromechanical Behavior of PZT Ceramic and Micro-Actuators

Jin Fan, Yue Donghua, Zhao Fengpeng, Shen Zhanpeng, Shi Pingan

2019, 40(2): 189-196. doi: 10.13832/j.jnpe.2019.02.0189

Abstract(208)

Abstract:
To explore the effects of He ion irradiation on electromechanical behavior of PZT ceramic and micro-actuators, He ion irradiation experiments with three different total doses have been performed on PZT ceramic. By combining nano-indentation experiments and finite element analysis, anisotropic elastic parameters of PZT materials have been identified. The corresponding piezoelectric and dielectric constants are obtained with use of electronic test instruments. Based on the material parameters, deformation response and natural frequencies of the micro-actuators are calculated by finite element analysis, and the resulting irradiation effects are assessed. With increasing magnitude of total doses, the elastic, piezoelectric and dielectric constants of PZT ceramic show a decrease to some extent, with decay rates less than 10%, 4.8% and 16.6%, respectively. Variation of the maximum deformation and tensile stress of micro-actuators is less than 4.6% and 5.4%, respectively. The natural frequencies of micro-actuators vary less than 1%.

Dynamic Response of Sandwich Structures with Pre-Formed Holes under Blast Loading

Li Wei, Qian Lixin, Feng Gaopeng, Lu Yonggang

2019, 40(2): 197-202. doi: 10.13832/j.jnpe.2019.02.0197

Abstract(223)

Abstract:
Considering the structure preforated by fragments, the dynamic response of the plane sandwich with pre-formed holes under blast loading was investigated by employing the numerical simulation. The numerical models based on structural parameters have been established. The structural dynamic response process has been divided into three specific stages and the deformation modes have been classified and discussed systematically. Based on this, the effect of the number of pre-formed holes on the structural dynamic response has been investigated. It is concluded that the radial kinetic energy of the pre-formed holes resulted from rarefaction wave obviously decreases the vertical displacement of the structure, and the increasing of the number of the pre-formed holes changes the 2D face bearing capacity to 1D beam bearing capacity, to increase the vertical displacement and obviously decrease the bearing capacity. The research results provide a foundation for the analysis of the dynamic structural response and vulnerability of the sandwich structure with pre-formed holes under blast loading, and also provide a reference for the design of sandwich structure protection.

Research on Euler-Lagrange Coupling Multi-Purpose Code for Explosion Dynamics Analysis

Pu Xifeng, Wang Zhongqi

2019, 40(2): 203-207. doi: 10.13832/j.jnpe.2019.02.0202

Abstract(276)

Abstract:
In this paper, a Coupled-Euler-Lagrange multi-purpose code for explosion dynamics analysis has been developed combining the characteristics of Eulerian algorithm and Lagrangian algorithm. The code contains several material models and equations of state to describe the behavior of the material under different types of loading. The test results show that with this Euler-Lagrange coupling code, the interaction between fluids and solids can be monitored and evaluated, and the loading and the interaction process between the flow field and the structure can be analyzed in a better way. At the same time, it can provide a better analysis platform for the structural safety evaluation under explosive and impact loads in nuclear engineering, and also provide a basic platform for the development of related multi-physical field coupling computing technology.

2019 Vol. 40, No. 2