核动力工程

Simulation Model Architecture and Concept Validation for Thermal Hydraulic Characteristics of Two-Phase Fluid Based on Modelica

Huang Yanping, Zeng Xiaokang, Ding Ji

2021, 42(1): 1-7. doi: 10.13832/j.jnpe.2021.01.0001

Abstract(1826) PDF(382)

Abstract:
Considering the latest improvement of international simulation technology, this paper studies the advanced simulation technology for nuclear reactor thermal-hydraulic system based on Modelica. The model architecture of thermal-hydraulic characteristics of the two-phase flow is established The layering architecture figure is constructed based on the basic control equations model and system model, and validated by the simulation of thermal-hydraulic characteristics of the two-phase flow in the pipe based on Modelica, to demonstrate its feasibility. The study shows that the hierarchical modeling technology based on Modelica can easily extract the generality and base class as basic models in the thermal-hydraulic characteristics of two phase flow of the nuclear system. By using multi-stage inheritance method, the Modelica model for specific object can be constructed by parameter configuration and association combination of basic level models, and the component model or system model can be formed by combination encapsulation or drag and drop modeling. The simulation result of the thermal-hydraulic characteristics of the two phase flow in the pipe by Modelica agrees with that by Relap5. The Modelica modeling approach is standard, efficient and flexible, which is suitable for the collaborative simulation of complex physical systems

Research on Density Wave Oscillations in Steam Generator Heat Transfer Tube

Liu Mengmeng, Zhang Zhen, Yang Xingtuan, Jiang Shengyao

2021, 42(1): 8-14. doi: 10.13832/j.jnpe.2021.01.0008

Abstract(415) PDF(149)

Abstract:
There exists two-phase flow in the heat transfer tube of the steam generator, in which the density wave oscillation (DWO) phenomenon plays a vital role in the safe operation of the nuclear reactor. This paper studies the DWO phenomenon in the tubes under bilateral convection by numerical calculation. Firstly, comparing the results of the numerical model with Babcock & Wilcox Once-Through Steam Generation (OTSG) experiment, the numerical model is verified. Secondly, the DWO boundary in different boundary heating conditions (bilateral convection and traditionally electric heating on the wall) is compared. Lastly, the movement of DWO boundary is analyzed by changing the height of the heating section and flow directions. It is found that increasing the length of the heating section and appropriately reducing the level inclination (within the range of 50 to 90 degree) of the tubes can expand the stable interval in the system. The main purpose of this paper is to provide reference and guidance for the safe operation of the steam generator.

Experimental Research on Flow Instability in Multi-Parallel Channels

Wang Yanlin, Zhou Lei, Zan Yuanfeng, Xu Jianjun, Yan Xiao

2021, 42(1): 15-17. doi: 10.13832/j.jnpe.2021.01.0015

Abstract(306)

Abstract:
The flow instability in the multi-parallel channels was investigated under the condition of: P=0.89~1.32 MPa, G=500~750 kg/ (m2?s), and ΔTsub,in=58.5℃~132.3℃ with the upflow deionized water. The test section consists of 2, 3 and 5 parallel tube channels with the dimension of 1400 mm×Φ8 mm×2 mm. The characteristics of the flow instability in the parallel channels and the flow instability boundary was studied based on the experimental results. It shows that the flow excursion appears when the water in the channels becomes two phase, and period oscillation appears when the mass quality at the exit of the channel reaches to a certain level. The number of the heated channels has little effect on the flow instability boundary.

Research on Flow Field in Rod Bundle Channel under Low Reynolds Number Using PIV Technique

Qi Peiyao, Deng Jian, Tan Sichao, Qiu Feng, Yu Xiaoyong

2021, 42(1): 18-22. doi: 10.13832/j.jnpe.2021.01.0018

Abstract(278)

Abstract:
Based on the particle image velocimetry (PIV) technique, the flow visualization in the fully developed 5×5 rod bundle channel at low Reynolds number was studied. The Reynolds number ranges from 310 to 12296. The experimental results show that the relative velocity gradient in the rod bundle channel is more significant at low Reynolds number. With the increasing of Reynolds number, the velocity distribution in the rod bundle channel tends to be uniform. The transition observed by the resistance characteristics in the rod bundle channel are more ambiguous than those in the circular tube, and the transition Reynolds number is about 900. Under the influence of low Reynolds number effect, the dimensionless RMS velocity decreases with the increasing of Reynolds number, while near the transition Reynolds number, the dimensionless RMS increases with the increasing of Reynolds number. Besides, the experimental data can be used to verify the applicability of the turbulence model to different Reynolds Numbers.

Theory Study on Effect of U-Tube Length on Reverse Flow in UTSG

Wang Yihu, Lu Chuan, Cong Tenglong, Chen Yiran, Xin Sufang, Huang Huijian, Xu Liangjian, Wu Yingwei

2021, 42(1): 23-27. doi: 10.13832/j.jnpe.2021.01.0023

Abstract(281)

Abstract:
Under the natural circulation of the primary loop, the backflow may occur in the partial U-tubes of an inverted U-tube steam generator (UTSG), which brings negative effect on natural circulation. The relationship between the critical pressure drop and the tube length of the inverted U-tube is deduced by theoretically analyzing the hydrodynamic curve of the inverted U-tube of a steam generator in a UTSG, and the numerical analysis is performed using the system analysis program RELAP5. It is shown that the critical pressure drop decreases first and then increases when the pipe length increases, that is, the shortest inverted U-shaped tube of the small UTSG steam generator first reverses, while the longest tube of the large UTSG steam generator first reverses. The conclusions explain the phenomenon of different distribution of the reflux tubes obtained by different simulations, and provide a theoretical basis for the prediction of the spatial distribution of the reflux tubes in the inverted U-tube of the UTSG.

Experimental Study of Diameter Effect on Gas-Liquid Two-Phase Countercurrent Flow Limitations in Vertical Pipes

Ma Youfu, Shao Jie, Lu Peng, Zhu Kui

2021, 42(1): 28-34. doi: 10.13832/j.jnpe.2021.01.0028

Abstract(161)

Abstract:
For the safety analysis of pressurized water reactors, it is necessary to accurately predict the gas-liquid flow relationship under the gas-liquid countercurrent limitation (CCFL) conditions. In this study, with the use of the experimental method of submerged exhaust, the CCFL characteristics of the vertical pipes with different pipe diameters and an identical pipe length were tested. Then, the CCFL correlation models for vertical pipes were analyzed. The main conclusions are: ①The flow pattern in vertical pipes is annular flow under the CCFL condition. When the superficial gas velocity is large, the liquid film in the large diameter pipe shows a churn-like motion, while the liquid film in the small diameter pipe behaves a stable falling flow with wavy interface. As the superficial gas velocity decreases, they both convert to a falling film flow with smooth interface. ②The model based on Wallis numbers presents an over-correlation to the effect of pipe diameters on the CCFL of vertical pipes; meanwhile, the models based on the Kutateladze number or the Froude-Ohnesorge number also fail to correlate well for the diameter effect on the CCFL of vertical pipes. ③A new dimensionless parameter is proposed to depict the CCFL of vertical pipes, which can correlate the diameter effect satisfactorily and can indicate the effect of fluid properties on the CCFL of vertical pipes simultaneously.

Experimental Study on Heat Transfer Characteristics of Subcooled Flow Boiling in a Narrow Vertical Rectangular Channel

Yan Tianyu, Wang Teng, Bi Qincheng, Wang Zehao

2021, 42(1): 35-41. doi: 10.13832/j.jnpe.2021.01.0035

Abstract(306)

Abstract:
Experimental investigation on the heat transfer characteristics of the deionized water flow was performed in a narrow vertical rectangular channel. The cross-sectional dimension of the flow is 50 mm×2 mm. The experiments were performed with pressures at the inlet of the test section ranges from 0.1 to 0.3 MPa, mass flux ranges from 200 to 1400 kg?m-2?s-1, and heat flux ranges from 20 to 320 kW?m-2. The boiling curves and the heat transfer curves of water were obtained in narrow rectangular channel. Compared with 8 empirical correlations separately, and adopting similar theories and regression analysis, a new correlation is proposed for the narrow vertical rectangular channel in the subcooled flow boiling region. The study results show that in the narrow vertical rectangular channel, the heat flux shows the major enhancement effect as subcooled flow boiling happens, the Bertsch correlation predicts the experimental data best among those empirical correlations, and the new correlation predicts the experimental data well. Thus, the new correlation used in the paper is applicable in the prediction of the heat transfer coefficient of subcooled flow boiling in the narrow vertical rectangular channel.

Applicability Analysis of Onset Model for Debris Bed Relocation in Sodium-Cooled Fast Reactor

Teng Chunming, Zhang Bin, Shan Jianqiang, Zhang Xisi, Cao Yonggang

2021, 42(1): 42-47. doi: 10.13832/j.jnpe.2021.01.0042

Abstract(178)

Abstract:
When the core disruptive accident (CDA) occurs in the sodium-cooled fast reactor (SFR), the melted core fuel is cooled and solidified into debris particles by the coolant, which accumulates in the lower plenum and forms the debris bed. In order to effectively remove the decay heat from the debris bed and implement the in-vessel corium retention (IVR), it is necessary to predict and simulate the relocation behavior of the debris bed. For this reason, Zhang et al. developed an onset model for judging the relocation of the debris bed based on the force balance analysis of the particle. In this paper, a large number of debris bed relocation experiments were carried out by bottom gas-injection method, and the relocation onset model was used to calculate and predict the relocation behavior of the debris bed under different experimental conditions, which further verified the applicability of the relocation onset model.

Numerical Simulation of Pipeline Erosion of Particulate Matter in LBE Based on DPM Model

Du Xiaochao, Liu Shuai, Liu Peng, Hong Feng, Yuan Xianbao, Zhang Yonghong

2021, 42(1): 48-53. doi: 10.13832/j.jnpe.2021.01.0048

Abstract(434)

Abstract:
In order to study the erosion of liquid lead-bismuth eutectic(LBE) alloy fluid with particles, the fluid erosion on the pipe wall was numerical simulated by the discrete phase model (DPM) of Fluent, based on the theory of fluid mechanics. The results show that the angle of the elbow, the particle size, the particle concentration, the pipe diameter and the velocity of the flow have a significant effect on the erosion wear of the pipe wall. At high flow rates, the erosion is serious, and the erosion on the straight pipe section is much weaker. The maximum erosion rate of the elbow is mainly distributed in the range of 30° to 90°. The erosion of micro particles in the range of 1μm and 9 μm is small, but cannot be ignored. When the particle size is in the range of 10μm to 90μm, the erosion rate is not much different from that of micro particles. When the particle size increases to the range of 100 μm to 900 μm, the erosion of large-diameter particles is severe.

Research on Shutdown Protection Scheme for Sodium Cooled Fast Reactor

Xu Weidong, Duan Tianying, Fu Hao, Feng Weiwei, Yang Peng

2021, 42(1): 54-60. doi: 10.13832/j.jnpe.2021.01.0054

Abstract(242)

Abstract:
Based on the modeling experience of CEFR thermal-hydraulics model, the main system model of CFR is established through Relap code. According to the Anticipated Transient without Scram (ATWS) in the safety analysis of the fast reactor, the safety margin and shutdown protection in the event of reactivity accident introduction are studied. The results show that when the accident occurred at rated power, the alarm and shutdown signal of short-period was not triggered. Besides, the current setting value of each protection parameters in the primary loop, the signal measurement delay and drop time of scram rod can be taken from other values, when the reactivity insertion accident of uncontrolled shim rod withdrawal for 5 s or 10 s occurs. While for 15 s, the power and P/F signal can ensure that the reactor status meets the requirements of the accident acceptance criteria in the current design. To achieve the same function for shutdown protection triggered by sodium temperature at core outlet, it is necessary to make sure that the reactor is in depth subcritical before 14.85 s when the other protection signals fail.

Validation and Evaluation of COSINE Based on Gen Ⅲ Passive PWR Low Power Physical Test

Zhang Weibin, Zhu Chenglin, Wang Xing, Wang Shiwei, Zheng Zheng, Li Shuo, Yu Hui

2021, 42(1): 61-64. doi: 10.13832/j.jnpe.2021.01.0061

Abstract(497)

Abstract:
Based on the low-power physics test results of Gen Ⅲ Passive PWR AP1000 reactor, the calculation function and calculation accuracy of the COSINE software package design software are confirmed and evaluated. From the comparison of the control rod worth, ARO (All Rods Out) end boron concentration and ARO isothermal temperature coefficient, it can be seen that the calculation results of COSINE software package nuclear design software are in good agreement with the measured data in the low power physical test of AP1000 reactor, and meet the engineering design requirements, with good calculation accuracy.

Investigation on Resonance Self-Shielding Calculation for PWR Cladding Materials Based on Equivalence Theory

Xiao Xiang, Wu Jun, Chen Yixue, Yang Tongrui, Zhu Chenglin, Li Shuo

2021, 42(1): 65-69. doi: 10.13832/j.jnpe.2021.01.0065

Abstract(286)

Abstract:
The treatment of resonance self-shielding effect is one of the key factors for reactivity accuracy in PWR lattice code. The PWR cladding materials for zirconium isotopes also has the resonance self-shielding effect. Ignoring the influence of resonance self-shielding effect for cladding materials, it can cause 100~300 pcm errors in reactivity predictions. Now, the reference dilution cross section and cladding equivalence theory can be used in treating the resonance self-shielding effect for cladding material, but the applicability and accuracy of these methods has not been completely tested. Therefore, a series of PWR cases are used in testing the applicability of these methods computed by DRAGON code. The main impact factors for the resonance self-shielding effect in cladding material and the applicability of these two methods are determined. The results show that the resonance self-shielding effect in cladding material is only related to nuclide density, thickness and water/uranium ratio in the moderator region. The reference dilution cross section satisfies the typical PWR cases, however it can cause more errors in the great change of the above three factors. The new equivalence theory is more accurate and flexible, which can be used in different PWR cladding materials for the calculation of the resonance self-shielding effect.

Analysis of Three-Dimensional Nonlinear Impact on Reactor Structure

Cao Guitao, Zhu He, Ye Xuexiang

2021, 42(1): 70-74. doi: 10.13832/j.jnpe.2021.01.0070

Abstract(235)

Abstract:
In order to analyze the dynamic response of the reactor structure, a three-dimensional nonlinear was established. The model overcame many nonlinear factors such as clearance, contact, friction, damping, pretension, impact, fluid-structure coupling and connection stiffness. Considering the fluid-structure coupling between the core barrel and the RPV, the hydrodynamic mass matrix was modeled and verified by ANSYS acoustic element. Three-dimensional upper core plate, lower core support plate was set up. Multi-group fuel assembly was built, which considered the preload and bounce with the core plate, the clearance and impact with the core shroud. The model accurately simulated the dynamic response of the reactor structure because of the coupling of horizontal and vertical directions. Finally, the response of each component was obtained by inputting the time history of the impact acceleration in three directions, which provided the input for the stress analysis of the components and the impact experiment of the control rod drive line. The method in this paper can offer a reference for the three-dimensional nonlinear dynamic analysis of the reactor structure.

Optimization of Main Heat Exchanger Design for Primary Circuit of Low Temperature Nuclear Heating Reactor

Zhang Yan, Ma Lanqing, Jin Dongjie, Zhao Binbin

2021, 42(1): 75-79. doi: 10.13832/j.jnpe.2021.01.0075

Abstract(177)

Abstract:
Optimization of the design of the general structure and flow path was carried out for the main heat exchanger of Low Temperature Nuclear Heating Reactor. The comparison of the schemes with and without optimization is described in detail. The structural stress and the external pressure buckling are analyzed in each working condition. The research shows that the main heat exchanger of the low-temperature nuclear heating reactor is with the advantages of simple structure, high economical performance, and easy of in-service inspection and manufacture, and thus is worth promoting and applying in such reactors.

Design of Periodic Test for Reactor Protection System of CENTER

Xiao Peng, Liu Hongchun, He Zhengxi, Zhao Yang, Li Wei, Tang Tao

2021, 42(1): 80-85. doi: 10.13832/j.jnpe.2021.01.0080

Abstract(215)

Abstract:
Reactor protection system of CENTER is based on NASPIC platform developed by NPIC. According to the requirement of GB/T 5204 and IEEE 338, based on the design criterion of periodic test, this paper introduces the overall scheme of periodic test for reactor protection system based on the principle of subsection and overlap as well as the characteristics of the NASPIC platform itself. T1 test, T2 test, T3 test, response time test and their principle are mainly described. This design can be referenced by other reactors.

Research on Pressurizer Pressure Control Loop Test Based on Minimum Verification Platform of DCS Transformation

Wang Juncheng, Wu Wenchao, An Tiancai, Liu Kaidi, Jiang Bo, Zhou Zhijiang, Zhu Meiying

2021, 42(1): 86-89. doi: 10.13832/j.jnpe.2021.01.0086

Abstract(145)

Abstract:
Daya Bay Nuclear Power Plant adopts U.S. Bailey9020 analog control platform and plans to carry out the digitally upgrading during 30-year overhaul. In this project, the target SH-N system, Bailey9020 platform and Daya Bay process simulation model are used to build the verification platform through the PLC interface, and the pressurizer pressure control loop is selected as the verification object. This project takes the startup test program as reference. By comparing the response differences of the same disturbance on different platforms, it verifies the applicability of the current control parameters of Daya Bay in the target DCS platform, and finds the inherent defect that the proportional function of the PI controller of bailey9020 platform cannot respond when conducting the manual/automatic switching. The corresponding optimization measures are put forward from the aspects of circuit design and operation program. This project achieves the purpose of analog board analysis, DCS configuration verification and constant value conversion, and provides important feedback for the current Daya Bay unit operation.

Research on Pressurizer Pressure Control Loop Test Based on Minimum Verification Platform of DCS Transformation

Xu Ying, Zhang Qiang, Xu Jinquan, Yu Hang

2021, 42(1): 90-94. doi: 10.13832/j.jnpe.2021.01.0090

Abstract(163)

Abstract:
Daya Bay Nuclear Power Plant adopts U.S. Bailey9020 analog control platform and plans to carry out the digitally upgrading during 30-year overhaul. In this project, the target SH-N system, Bailey9020 platform and Daya Bay process simulation model are used to build the verification platform through the PLC interface, and the pressurizer pressure control loop is selected as the verification object. This project takes the startup test program as reference. By comparing the response differences of the same disturbance on different platforms, it verifies the applicability of the current control parameters of Daya Bay in the target DCS platform, and finds the inherent defect that the proportional function of the PI controller of bailey9020 platform cannot respond when conducting the manual/automatic switching. The corresponding optimization measures are put forward from the aspects of circuit design and operation program. This project achieves the purpose of analog board analysis, DCS configuration verification and constant value conversion, and provides important feedback for the current Daya Bay unit operation.

A Comparative Study of Different Methodologies for Fast Estimation of External Gamma Radiation Doses inside Compartments

Yang Zhangcan, He Yingzhao, Wang Shuai

2021, 42(1): 95-99. doi: 10.13832/j.jnpe.2021.01.0095

Abstract(162)

Abstract:
For nuclear-powered ships, the fast estimation of external gamma radiation dose is important for accident consequence evaluation and emergency response plan. However, currently there is no standard for the fast estimation of external gamma radiation dose in compartments. In this study, the results by MCNP calculations are used as a baseline to evaluate the accuracy of two methodologies, namely the finite cloud immersion method and the point-kernel integral method. Comparisons are performed for compartments with different shapes and volumes. The results show that the errors become larger if the shape of a compartment is more different from a hemisphere, or the volume of a compartment is larger; the error of the finite cloud immersion method is about 30%, while the error of the point-kernel integral method is about 10%. Therefore, the point-kernel integral method can provide more accurate and fast estimation for practical applications.

Integration Technology of Functional Safety and Cyber Security for Nuclear Safety Class DCS

Jin Jianghong, Xia Qiaoli, Mo Changyu

2021, 42(1): 100-106. doi: 10.13832/j.jnpe.2021.01.0100

Abstract(211) PDF(83)

Abstract:
Taking the typical nuclear safety class DCS as an example, the FMVEA technology is used to evaluate the compatibility between functional safety and cyber security. The combination of event tree and risk analysis is used to give functional safety and cyber security coordination solutions, and finally obtain the integrated protective measures of nuclear safety class DCS functional safety and cyber security. The results show that the application of trade-off technology to obtain the cyber security measures can be better compatible with the original functional safety measures. Therefore, the trade-off techniques established in this study can be applied to nuclear safety class DCS cyber security design work.

Research on Coupling Simulation Using Simulink and CATIA

Shen Mengsi, Li Yiliang, Lin Meng, Xiao Kai

2021, 42(1): 107-111. doi: 10.13832/j.jnpe.2021.01.0107

Abstract(255)

Abstract:
The control system is coded in the transient analysis code CATIA, thus the logical of the control system cannot be optimized, which limits the application range of the code. Hence, the graphical modeling tool Simulink is used to model the control system, and the coupling simulation using the Simulink code and the system analysis code CATIA is finished. The computational results showed that the coupling simulation results agree well with the computational results using the system analysis code CATIA, but the modeling process of the control system with Simulink code is visual and with high efficiency. Meanwhile, the coupling method developed is extensible.

Research on PR Control for Rod Position Detector Power Supply in Nuclear Power Plant Based on Root-Locus Method

Wang Chunlei, Chen Shuaijun, Huang Kedong, Xu Mingzhou, He Jiaji, Li Mengshu, Li Guoyong, Zheng Gao

2021, 42(1): 113-117. doi: 10.13832/j.jnpe.2021.01.0113

Abstract(157)

Abstract:
According to the application requirements of the single phase inverter in the rod position detection system of nuclear power plant, the control strategy is analyzed based on the mathematical mode, and the realization mode of the output voltage outer loop proportional resonant(PR) controller in the discrete domain is mainly analyzed. The Matlab simulation and experiments show that the control method can keep the output voltage stable, good sinusity, fast dynamic response and strong anti-interference ability and so on, which is beneficial to improve the accuracy of rod position detection.

HFETR Initiating Events Analysis in Level 1 PSA

Zhou Chunlin, Wang Wenlong, Li Haitao, Zhang Jiangyun, Zheng Daji, Wu Wei, Deng Yunli, Liu Peng, Wei Fu

2021, 42(1): 118-122. doi: 10.13832/j.jnpe.2021.01.0118

Abstract(284)

Abstract:
In the level one Probabilistic Safety Analysis (PSA) of High Flux Engineering Test Reactor (HFETR), the primary task is the initiating event analysis. In the present paper, the methods such as engineering evaluation, reference to the previous list of initiating events, deductive analysis and operational experience summary are applied comprehensively to determine the list of PSA initiating events in the HFETR operation period. Then, the initiating events are properly grouped. Combined with the methods such as fault tree analysis, HFETR operation event statistics and reference to the same type of research reactor, the frequency of each initiating event group is given, which lays a foundation for the subsequent level 1 PSA of HFETR.

Research on Design of Pressurizer Degassing System for Small Module Reactors

Cai Zhiyun, Ren Yun, Lai Jianyong, Zhang Yulong, Liu Xianghong

2021, 42(1): 123-128. doi: 10.13832/j.jnpe.2021.01.0123

Abstract(206)

Abstract:
In order to overcome the problems of time-consuming and complex operation caused by the traditional degassing method of chemical and volume control system, a design scheme of pressurizer degassing system using the pressurizer for thermal degassing is proposed. A dedicated computer program is developed based on pressurizer steady degassing model and optimization algorithm, by which the pressurizer degassing optimization calculation and analysis for the small module reactor is carried out in full range of shutdown operation conditions, and the size of the optimized flow limiting orifice and influence factors are obtained. The evaluation with the practical degassing effects shows that, the degassing system proposed in this paper can complete the hydrogen removing task in 4.13 hours theoretically, which is with obvious advantages over the traditional degassing method of the chemical and volume control system.

Experimental Research on Pump Characteristics under Two-Phase Condition

Su Qianhua, Wang Kuo, Xing Jun, Hong Rongkun, Peng Fan, Lu Donghua

2021, 42(1): 129-132. doi: 10.13832/j.jnpe.2021.01.0129

Abstract(195)

Abstract:
For indigenous mega-kilowatt class reactor cooling pumps (RCPs) project, the performance characteristics of subject pump based on the scale model of RCPs are fully researched in the specialized experiment system. The pump characteristics under single-phase liquid condition in normal speed case, turbine case, energy dissipation case, rotor locked case and runaway speed case are firstly obtained. Furthermore, the pump two-phase specific operating conditions under the void fraction of 0.1, 0.2, 0.3, 0.4, 0.5, 0.6, 0.7, 0.8, 0.9 and 1.0 are studied one by one. Finally, we use the proportion law to sort out the experimental data under multiple pump speeds, and then obtain the change rule of the pump head under different void fractions and different flows. The curve of pump head proportional law on the basis of single-phase liquid has a process of continuous drift and regression with the increase of the void fraction. When the void fraction reaches about 0.7, the drift degree is the largest, and when the fluid changes to single-phase gas, the regression is realized. The above experimental results provide data support for further research of nuclear main pump design and primary circuit safety analysis.

Evaluation Method for CRDM Roller State Based on Evaluation Function and BP Network

Jiao Meng, Cai Qi, Zhang Liming, Yang Xiaochen, Zhang Yongfa

2021, 42(1): 133-137. doi: 10.13832/j.jnpe.2021.01.0133

Abstract(204)

Abstract:
In this paper, an evaluation method for CRDM roller state based on the evaluation function and error back propagation training (BP) network is proposed for the non-stationary and strong noise distortion signals in the control rod drive mechanism (CRDM) vibration signals. The signal is denoised by semi-soft threshold, the feature vectors are extracted by local mean decomposition (LMD), and the sample set composed of feature vectors is identified by BP network for state recognition. An evaluation function is introduced to evaluate the results of state recognition. The distorted samples are removed according to the evaluation results, and the new sample set is retained for state recognition. The results show that this method can effectively identify the defect state of the rollers and effectively solve the difficulty in accurate identification of the rollers state of the control rod drive mechanism.

Failure Analysis and Reliability Improvement of Control System of Main Steam Isolation Valve in Nuclear Power Plants

Li Xiaoquan

2021, 42(1): 138-142. doi: 10.13832/j.jnpe.2021.01.0138

Abstract(210)

Abstract:
The main steam isolation valve is the most important isolation equipment of the steam pipeline between the nuclear island and the conventional island in nuclear power plants. This paper mainly introduced the function and control theory, and the failure mode analysis. Combined with the historical fault statistics, the weak point of the system-limit switch is obtained, and the failure mechanism and root cause analysis of limit switch failure are emphatically expounded. At last, the paper puts forward the improvement measures from five aspects: human, machine, material, method and environment, which has certain reference significance for the maintenance and improvement of the main steam isolation control system of other nuclear power plants in service in China.

Quantitative Analysis and Study of β Radiation Damage to Instruments in Severe Nuclear Power Plant Accidents

Qin Yue, Zhu Jialiang, Zhang Mi, He Peng, He Zhengxi, Li Xiaofen, Xu Tao, Chen Jing, Li Hongxia, Ye Yuheng

2021, 42(1): 143-147. doi: 10.13832/j.jnpe.2021.01.0143

Abstract(204)

Abstract:
Based on the research and analysis of the β radiation distribution characteristics in the nuclear power plant within a severe accident, the radiation distribution is calculated when β radiation works on the typical instruments and when. γ radiation works on the same typical instruments. The β radiation harm influence is converted into a certain percentage of γ radiation through the way of energy equivalence, to obtain the β radiation damage quantitative data and the β radiation shielding performance data of different materials. At the same time, it provided the quantitative test radiation value in severe accident qualification. The research result can be used in engineering practice directly for conducting the instrumentation survivability analysis, equipment qualification and equipment design.

Investigation on O-Ring Failure of RCP Shaft Seal of 1080 MW Nuclear Power Plant and Its Optimization

Chen Kan, Liu Wei, Guo Yi, Ren Hebin, Zhang Junkai, Liu Qiang

2021, 42(1): 148-153. doi: 10.13832/j.jnpe.2021.01.0148

Abstract(142)

Abstract:
The performance of RCP hydrodynamic shaft seals in both three positions degraded near the end of operation cycle, which is the feedback from NO.103 overhaul in one Chinese domestic nuclear power plant. The low-pressure leakage is above normal level, and the pressure before No.3 seal face decreases by 15%, according to the field monitoring data. In order to find out the reason of this phenomenon, a full-scale test rig was set up. An optimal design of the secondary seal is applied by Optimization the dimension of the O-ring. The friction forces of the secondary seal couple are cross measured. The No.3 guide ring is updated by coating with Cr3C2. The results show that the friction force of the original secondary seal pair is above 300 N, and the friction force of the new design is reduced to less than 100 N. The seal leakage was reduced by 11.7%. After that, a total 500 hours’ endurable experiment is done. The sealing performance data and No.3 guide ring surface roughness value change were recorded before and after the experiment. The test data proves that the optimized design can improve the RCP shaft seal reliability.

Research on Modeling and Simulation of Once-Through Steam Generator

Xu Yu, HuangFu Zeyu, Xu Jianqun, Tian Peiyu, Li Chunmei, Cheng Xiang, Yan Siwei, Liao Xianwei

2021, 42(1): 154-160. doi: 10.13832/j.jnpe.2021.01.0154

Abstract(514)

Abstract:
Taking a once-through steam generator designed by Babcock & Wilcox as the study object, a graphical model was designed, based on the basic module of supporting platform APROS. Based on the simulation with the different two-phase flow calculation model, after comparative analysis, a six-equation model was selected for calculation. The simulation results from different working conditions are compared with the results of empirical formula for pressure and heat transfer calculation, and the fluid pressure and heat transfer efficiency in the model are modified. Then the steady-state and dynamic simulation results under the set working conditions are compared with the experimental data published abroad to verify the accuracy of the model.

Research on Positioning of HPR1000 Reactor Detector Replacement

An Yanbo, Yu Zhiwei, Li Na, Wang Bingyan, Xiong Siyong, Zhang Anrui

2021, 42(1): 161-166. doi: 10.13832/j.jnpe.2021.01.0161

Abstract(203)

Abstract:
In the process of replacing HPR1000 reactor detector, the first step is to position the reactor detector replacement above the detector. This paper proposes the positioning requirements of the reactor detector replacement firstly, and then analyzes the position system combined with the process of replacing reactor detector, and proposes the closed loop visual servo positioning scheme of reactor detector replacement. This technical scheme fulfills the position accuracy requirement, ensures the detector a small deflection angle during lifting and mitigates the wear of the mechanical seal joint .

Risks Analysis and Countermeasures in Natural Circulation Test of CAP1400 Passive Residual Heat Removal System

Li Wenkai, Li Ping, Li Wenshuang, Kong Xiangwei, Huang Yong

2021, 42(1): 167-171. doi: 10.13832/j.jnpe.2021.01.0167

Abstract(198)

Abstract:
The natural circulation test of CAP1400 passive residual heat removal heat exchanger system（PRHR）is one of the first plant only tests and the major plant transient tests during the commissioning. During the natural circulation test, the temperature, pressure, liquid level and other parameters of the reactor primary circuit change dramatically, which increase the test risk and propose higher requirements for the operation control of the unit. Based on the AP1000 commissioning test practice, it is proposed to perform the PRHR natural circulation test using real decay heat generated following the reactor operation from the perspective of reducing test risk. At the same time, it analyzes the safety risks caused by the drastic changes in the parameters of the primary loop pressure, temperature, pressurizer level and the external source range detectors of the reactor during the test, and formulates corresponding countermeasures to provide the strong support for the safety implementation of subsequent CAP1400 PRHR natural tests.

Study on Dynamic Adjustment Method of Preventive Maintenance Strategy for Affected Equipment under Long-Term Temporary Outage of Nuclear Power Units

Zhang Jianghong, Peng Buhu, Liu Xiaolei, Li Guomeng, Liu Haixiao

2021, 42(1): 172-176. doi: 10.13832/j.jnpe.2021.01.0172

Abstract(161)

Abstract:
This paper proposes a method for the dynamic adjustment of preventive maintenance strategies for equipments affected by unit temporary outage in view of the current situation of domestic nuclear power plants and the degradation of equipment reliability caused by temporary outage. This paper studies the identification method of equipment list affected by temporary shutdown, classifies the affected equipment and formulates the equipment maintenance strategy template. Finally, it optimizes and adjusts the maintenance strategy template to implement the application, and analyzes and explains the specific case. The results show that the adjustment of maintenance strategies for equipments affected by unit temporary outage can increase the operating reliability and maintenance effectiveness, which lays the foundation for the equipment reliability analysis during the long-term downtime of nuclear power plant units.

Study and Design of Rod Position Detector Excitation Power Based on Feedforward Control

Li Mengshu, Li Guoyong, Zheng Gao, Huang Kedong, Xu Mingzhou, He Jiaji, Luo Qiurong

2021, 42(1): 177-181. doi: 10.13832/j.jnpe.2021.01.0177

Abstract(157)

Abstract:
This paper presents a design scheme of rod position detector excitation power based on feedforward control. High-frequency switching AC-DC-AC inverter circuit is used in this scheme, and the feedforward control is added on the traditional Proportional-Integral (PI) control to suppress the influence of measurable and uncontrollable factors on the controlled object. Through theoretical analysis and simulation test, the design scheme proposed in this paper is verified, and it can reduce the output ripple, and improve the current amplitude accuracy while ensuring the corresponding control time.

Research and Development of Integrated Waste Resin Interface Measurement System

Liu Chenlong, Li Zhenchen, Zhang Yongkang, Xiong Wei, Dai Jun, Bai Ze, Xu Jie, Qiu Fumao, Zhou Chunyan, Bai Siyu

2021, 42(1): 182-185. doi: 10.13832/j.jnpe.2021.01.0182

Abstract(133)

Abstract:
At present, there is no instrument for waste resin or water interface measurement at home and abroad. Based on the magnetostrictive principle, a measurement method of waste resin / water interface using driving device as auxiliary means is proposed. On this basis, the overall design of the measurement system is carried out, and the corresponding measurement system is designed. The designed system is verified by experiments. The test results show that the measurement system realizes the measurement of waste resin or water interface, and is with the characteristics of high measurement accuracy, miniaturization, modularization, adjustable lifting speed, and easy installation and maintenance. The successful development of the system solves the problem of waste resin or water interface measurement, and it is worthy of promotion and further engineering application in the process of radioactive waste liquid treatment.

Experimental Study on Boiling Heat Transfer of Zr-4 Quenched in Water

Wang Zefeng, Deng Jian, Wang Jiageng, Zhang Yong, Liu Yu, Xiong Jinbiao

2021, 42(1): 186-191. doi: 10.13832/j.jnpe.2021.01.0186

Abstract(313)

Abstract:
In order to study the quenching behavior and heat transfer performance of Zr-4 rodlet in water, the visualization method is employed, and the temperature history of Zr-4 rodlet is measured during the quenching process in the present study. Based on one-dimensional inverse heat conduction code, the surface heat flux and the temperature of Zr-4 rodlet are calculated, and hence the boiling curve is determined. Zr-4 experienced the following 4 stages, including film boiling, transition boiling, nucleate boiling and natural convective boiling. The effect of axial distance and liquid subcooling on the quenching behavior and the boiling heat transfer is analyzed. It is found that as the liquid subcooling increases, the quenching time is decreased, and the minimum film boiling temperature is increased. Noticeably, the transition boiling and the nucleate boiling are sensitive to the local surface conditions. A correlation for minimum film boiling temperature is developed, which is significant for reactor safety.

Research on Flow Distribution Characteristics of Inverted U- Shaped Tube of Steam Generator under Natural Circulation Condition

Zhao Pengcheng, Yi Feng, Yu Hongxing, Shi Wei, Wang Tianshi, Xia Bangyang, Chen Baowen

2021, 42(1): 192-197. doi: 10.13832/j.jnpe.2021.01.0192

Abstract(428)

Abstract:
Taking the steam generator of the primary circuit system of the PACTEL PWR thermal experiment device as the research object, the pressure drop and the relationship between heat transfer and fluid flow of the U-tube are established based on the equations of mass conservation, momentum conservation and energy conservation of the one-dimensional fluid flow model. Then, the flow distribution calculation program of the inverted U-tube steam generator based on genetic algorithm was developed, and the correctness and reliability of the program were verified by benchmark experiments. Finally, the flow distribution among the U-tube groups of the steam generator is calculated by using the flow distribution program, and the influences of the tube height, tube length and the heat transfer coefficient of the first/second side on the flow distribution within the steam generator are studied. The results of the developed flow allocator are in good agreement with the experiment. Under the selected natural cycle conditions, the backflow is more likely to occur in the long tube of the steam generator, and the backflow phenomenon is characterized by a wide distribution range and low single-tube flow rate. The velocity of the forward flow in an inverted U-tube is inversely proportional to the length of the tube and proportional to the height of the tube. The backflow velocity remained unchanged with the increase of pipe length and was inversely proportional to pipe height. When the coefficient of heat transfer is low, the total flow is inversely proportional to the heat transfer coefficient . When the heat transfer coefficient is higher than a specific value, the backflow occurs in part of the tubes and the total flow drops sharply.

Study on Important Phenomenon Identification and Numerical Simulation Uncertainty Analysis for PWR Large Break LOCA

Zeng Wei, Wang Jie, Huang Tao, Chen Wei, Ding Shuhua, Deng Chengcheng, Yang Jun

2021, 42(1): 198-203. doi: 10.13832/j.jnpe.2021.01.0198

Abstract(340)

Abstract:
Large break loss of coolant accident is one of the most important design basis accidents, the accurate calculation can provide great support to enhance the plant power. In this study, the best estimate program RELAP5 is used to simulate the condition FP-LP-2 during the experiments of the loss of fluid test (LOFT), and the Gesellschaft für Anlagen-und Reaktorsicherheit (GRS) uncertainty analysis method is applied to the uncertainty quantification and sensitivity analysis. The uncertainty envelope with 95% confidence of key output parameters is given, and the uncertainty trend of these output results and their reasons are analyzed. Sensitivity analysis shows that the factors with effects on the peak cladding temperature include core decay heat, the critical flow discharge coefficient of the intact loop break, and the thermal conductivity of the fuel rod. This study confirmed the effectiveness of the GRS method, and can provide support to improve the safety analysis method.

Resonance Calculation Method for Non-Rod-Type Fuel in NECP-X Based on Global-Local Coupling Method

Cao Lu, Liu Zhouyu, Zhang Minwan, He Qingming, Wen Xingjian, Cao Liangzhi, Ji Wenhao

2021, 42(1): 204-210. doi: 10.13832/j.jnpe.2021.01.0204

Abstract(238) PDF(45)

Abstract:

Based on the global-local coupling resonance calculation method that has been developed in the NECP-X program, this paper studies the resonance calculation method for non-rod-type fuels. Firstly, the Dancoff correction factor of the real problem is calculated by the neutron current method to deal with the global spatial effect. Then, the pitch of the moderator around small-scale problem is obtained based on the equivalent Dancoff correction factor. Finally, for the calculation of the effective self-shielding cross-sections of the small-scale problem, the resonance pseudo-nuclides subgroup method is used. The method is applied to the numerical calculation of non-rod-type fuel. Numerical results show that when compared with the Monte Carlo results, the maximum microscopic absorption cross-section deviation is less than 1.8%, and the k_inf deviations do not exceed 110 pcm when the method deals with the resonance of non-rod-type fuel pins. These results show that the method has high calculation accuracy; in the calculation of large-scale problem, the deviation of keff of the JRR-3M which is based on plate fuels is about 300 pcm during its burnup life, and the maximum assembly power deviation is less than 0.62%. Therefore, the resonance calculation method proposed in this paper has a high accuracy and precision.

Development and Verification of Direct Transport Code KuaFu

Zhao Chen, Peng Xingjie, Zhao Wenbo, Yu Yingrui, Li Qing

2021, 42(1): 211-216. doi: 10.13832/j.jnpe.2021.01.0211

Abstract(122)

Abstract:
To develop the adaptability of the direct transport method for advanced reactors with complicated geometry, the constructive solid geometry method and the 2D/1D transport method were adopted in a newly-developed direct transport code KuaFu. C++/Python hybrid programming was used. A 2D/1D transport solver with CMFD acceleration method and parallel technology was developed in KuaFu. With the validation of the C5G7 benchmark, these techniques were evaluated, and the difference between KuaFu and Monte-Carlo code MCNP are listed. All results show that KuaFu is with good accuracy compared with reference results. The CMFD and parallel method improves the efficiency significantly.

A Scheme for Solving SN Equation on Optical Thick and Strong Anisotropic Scattering Medium

Li Zhipeng, Cheng Tangpei, Yang Chao

2021, 42(1): 217-223. doi: 10.13832/j.jnpe.2021.01.0217

Abstract(130)

Abstract:
In order to realize fast calculation for the reactor facility shielding problem, it is necessary to establish a spatial discrete scheme for SN equation that can reduce the negative numerical flux, ensure the linearity of the iterative scheme, and achieve good accuracy in a large grid with minor calculation effort. This study is based on the spatial analytic solution of the within-group fixed source SN transport equation, and uses the analytic basis function method to expand the spatial term. In order to improve the calculation efficiency, the weighted coefficient method is used to avoid a large number of exponential operations in a single grid. High order source expansion is studied to improve the precision. Finally, a fast iterative scheme for the self-scattering source iteration is realized based on the Krylov subspace method. The numerical results show that this method can obtain great advantages in the optical thick medium with uniform material area, and can be used for the rapid calculation of shielding problem of large reactor building.

2021 Vol. 42, No. 1