核动力工程

Research of Optimization Technology for Equilibrium Cycle with Gadolinium

Ma Zirong, Su Jian, Zhou Sheng

2021, 42(3): 1-5. doi: 10.13832/j.jnpe.2021.03.0001

Abstract(279) PDF(97)

Abstract:
The design optimization of gadolinium rods is determined by analyzing the fuel management technology used for equilibrium cycle, and the optimization methods are to raise the bearing enrichment of the UO₂-Gd₂O₃ pellets and loading fully enriched UO₂ pellets at both ends of the gadolinium rods. The length of fully enriched UO₂ pellets at both ends of the gadolinium rods is determined by analyzing the effects of axial power distribution for different lengths. The bearing enrichment of the UO₂-Gd₂O₃ pellets is also determined by analyzing the effects for fuel economy and fuel manufacture. The effects of core power distribution when raising the bearing enrichment of the UO₂-Gd₂O₃ pellets and loading fully enriched UO₂ pellets at both ends of the gadolinium rods are analyzed individually and systematically. The effects of safety in condition Ⅰ and Ⅱ for design optimization are analyzed. Safety for the UO₂-Gd₂O₃ pellets in optimization case has been confirmed. It is found that the axial power distribution can be improved by loading fully enriched UO₂ pellets at both ends of the gadolinium rods, and the maximum local linear power density of UO₂-Gd₂O₃ pellets under reactivity accidents in condition Ⅱ can be reduced. UO₂-Gd₂O₃ pellets is within the melting limit and the safety margin can be improved by about 5.6% in safety analysis. The fuel cost can be saved by about 23 million yuan per cycle for each core with gadolinium. Thus, the optimization method of gadolinium rods design can be used to improve the fuel management for the operating units.

Development and Verification of Fast Reactor Multi-Group Cross Section Database Processing Code MGGC1.0

Huang Zifeng, Ma Xubo, Zhu Runze, Li Yaozhou, Zhang Bin

2021, 42(3): 6-13. doi: 10.13832/j.jnpe.2021.03.0006

Abstract(363) PDF(72)

Abstract:
为产生高精度的快堆截面数据，基于一致性N阶的勒让德函数（P_N）近似方法与临界曲率搜索方法，开发了快堆多群截面处理程序MGGC1.0，并进行了多方面基准验证。通过对均匀混合介质的宏观截面验证表明，中子产生截面的相对偏差均小于0.1%，裂变能谱的相对偏差均小于0.25%，总截面由于修正方式不同导致偏差稍大，但绝大多数能群的相对偏差都在0.5%以内。在临界基准实验中与蒙特卡罗程序RMC采用连续点截面的计算结果相比，78%的基准题的偏差都在100 pcm（1 pcm=10^-5）以内，表明MGGC1.0处理截面的精度较好。在此基础上，采用钠冷快堆基准题BN-600进行计算，与基准题参考计算结果相比，输运与扩散2种方法计算所得有效增殖因子的相对偏差分别为0.112%和0.09%，燃料多普勒系数和燃料密度系数的相对偏差分别为1.49%和1.37%，而结构材料钢的多普勒系数与密度系数的相对偏差稍大，分别为18.75%和24.31%，初步分析，偏差较大的原因与窄共振近似的处理方法有关。对于区域的功率分布，基于局部能量沉积模型计算得出的区域功率分布分数与基准参考解的偏差在0.3%之内，符合较好。

Development and Validation of Integrate CHF Correlation Development System

Liu Wei, Li Zhigang, Lu Qi, Du Sijia, Liu Yu, Deng Jian, Hu Ying

2021, 42(3): 14-17. doi: 10.13832/j.jnpe.2021.03.0014

Abstract(325) PDF(57)

Abstract:
In the manual development of rod bundle CHF correlation, there are shortcomings and deficiencies such as numerous links, tedious process, large amount of data, high error rate and long computing time, so the integrate CHF correlation development system (ICODES) is researched and implemented by Nuclear Power Institute of China (NPIC). This paper introduces the theory and structure of the ICODES, and performs the validation based on the rod bundle CHF test data and the basic shape of homemade CF-DRW correlation from NPIC. This study indicates that ICODES satisfies the requirements of the development of CHF correlation for fuel assembly design.
Key words: CHF correlation, Development, Validation

Prediction of Natural Convection Characteristics of Spent Fuel Vessel with Multi-Fluid Domains and High Temperature Difference Based on Non-Boussinesq Model

Long Teng, Zhang Guihe, Gao Chen, Liu Pan, Deng Xiaoyun, Jin Ting, Xiong Guangming

2021, 42(3): 18-25. doi: 10.13832/j.jnpe.2021.03.0018

Abstract(268) PDF(56)

Abstract:
Non-Boussinesq model is established based on Boussinesq model and full buoyancy model, to simulate the natural convection flow in multi-fluid domains with high temperature difference. The comparison between predicted profiles and experimental data in a single-fluid domain and in multi-fluid domains with high temperature difference is conducted by implementing respectively Boussinesq model, full buoyancy model and Non-Boussinesq model. The results show that for single-fluid domain, the predicted velocity and temperature using Non-Boussinesq model and full buoyancy model coincide with the experimental data with average derivation less than 9%. Boussinesq model is less accurate in simulating the velocity in high temperature different region. For multi-fluid domains, accurate velocity profiles in all fluid domains are not guaranteed for the full buoyancy mode. However, developed Non-Boussinesq model can not only simulate precisely the buoyancy force caused by density variation, but also guarantee an acceptable velocity results for multi-fluid domains. The momentum source term analysis and CFD analysis in multi-fluid domains coincide well with each other. The buoyancy force model is with significant effect on the velocity profile in multi-fluid domain problem, but with little effect on temperature profile prediction. Hence, the prediction method used in this paper can be used in the natural convection prediction in multi-fluid domains with high temperature difference.

Experimental and Numerical Investigation of Cavitation-Induced Choked Flow in Venturi Tube

Zhou Junjie, Song Yuchen, Wang Dezhong, Yin Junlian

2021, 42(3): 26-33. doi: 10.13832/j.jnpe.2021.03.0025

Abstract(518) PDF(117)

Abstract:
When an over-flow condition occurs in a PWR nuclear power plant, the venturi flowmeter downstream of the make-up pump is required to form a cavitation-induced choked flow to protect the flowrate from exceeding the limit. Using the method of FLUENT numerical simulation and high-speed camera experiments, three different cavitation models were used to study the cavitation-induced choked flow phenomenon, cavitation development law and flow characteristics of the venturi tube. The results show that the ZGB cavitation model and the SST k-ω turbulence model can be used to simulate the cavitation flow limitation phenomenon of the venturi tube accurately. During the flow limitation, a periodic cavitation phenomenon will occur inside the Venturi tube. Under the action of the wall re-entrant jet, micro flow behaviors, such as small bubbles falling off, tail bubbles falling off and the whole vapor cavities falling off, will happen.

Numerical Simulation of Leakage Characteristics of Pressurized Water Reactor Nuclear Power Plant

Yin Songtao, Wang Ningning, Wang Haijun, Zhu Mengxin

2021, 42(3): 32-37. doi: 10.13832/j.jnpe.2021.03.0032

Abstract(294) PDF(71)

Abstract:
This paper aims to propose a two-phase critical flow model to accurately evaluate the leakage rate of pipeline breaks. The critical flow model considering the nucleation of metastable liquids and non-equilibrium mass transfer is based on a two-fluid model coupled with an isothermal bubble growth model. The model is implemented by an explicit difference algorithm. The proposed model calculates fast and the model predictions exhibit strong similarities with the experimental data. The numerical research of the leakage process for critical flow is conducted. The results show that the subcooling has a significant effect on the fluid flow and the mass transfer process, while the upstream pressure only affects the fluid flow. The proposed model can provide a theoretical basis for the safety analysis of pressurized pipelines and vessels.

Investigation on Reversed Flow Characteristics of U-Tube Steam Generator under Ocean Conditions

He Gening, Li Xiaojia, Cong Tenglong, Chen Yiran, Li Donghui, Wu Ge

2021, 42(3): 37-42. doi: 10.13832/j.jnpe.2021.03.0037

Abstract(117) PDF(42)

Abstract:
Based on RELAP5 program, a dynamic simulation model of U-tube steam generator (SG) under typical ocean conditions is established and verified. In the reactor start-up period, the effect of different rolling conditions on the flow in the U-tube of SG under low flow forced cycle condition is studied. The results show that the reversed flow can be suppressed by delaying the introduction of ocean conditions. It is easier to reverse flow under rolling motion condition around the axis that parallel to the axis of SG tube bundle elbow.

Investigation on Chaotic Evolution of Natural Circulation Pressure Drop Oscillation under Rolling Condition Based on Maximal Lyapunov Exponent

Lin Yuqi, Gao Puzhen, Li Zongyang, Zhang Yinxing, Wang Zhongyi

2021, 42(3): 42-48. doi: 10.13832/j.jnpe.2021.03.0042

Abstract(274) PDF(31)

Abstract:
In this paper, a series of experiments on the chaotic evolution of pressure drop oscillations (PDO) under rolling motion were designed and performed. The chaotic characteristic is analyzed based on maximal Lyapunov exponent (λ). It is concluded that the chaotic evolution route can be divided into 3 regions. In the thermal-hydraulic stable region, it is a stochastic low-amplitude vibration and PDO cannot happen. In the frequency lock oscillation region, PDO is forced to keep the same frequency as rolling motion and forms quasi-periodical oscillation with increasing λ. In the rolling effect weakening region, PDO with natural frequency can generate reverse flow. λ reaches the maximum value and shows strong chaotic characteristics.

Study on Effect of Anisotropic Scattering Cross Section on Sensitivity Coefficient Calculation for Fast Reactors

Wang Dongyong, Ma Xubo, Zhu Runze, Zhang Bin, Peng Xingjie, Wang Lianjie

2021, 42(3): 48-55. doi: 10.13832/j.jnpe.2021.03.0048

Abstract(246) PDF(46)

Abstract:
Due to the high energy of neutrons in fast-spectrum reactors, the anisotropic scattering of neutrons is with great effect on the calculation results. In this study, the effects of high-order scattering cross-section perturbation on the calculation of elastic and inelastic cross-section sensitivity coefficients are studied when calculating the elastic and inelastic cross-section sensitivity coefficients. The causes of implicit sensitivities and related approximate conditions were theoretically analyzed. The direct perturbation method was used to calculate the sensitivity coefficients of the main nuclides reaction channel of the ZPR-6/7 fast-spectrum reactor. The research results show that for the ZPR-6/7 fast-spectrum reactor, without disturbing the ²³⁸U high-order scattering cross section, the sensitivity coefficient of the total elastic scattering cross section is 44.3% higher than that when the high-order scattering cross section is considered. Irrespective of the disturbance of ⁵⁶Fe high-order inelastic scattering cross section, it will cause the sensitivity coefficient of inelastic scattering cross section to be 28.9% higher, but it has less influence on the sensitivity coefficients of elastic scattering and inelastic scattering of other nuclides. After considering the high-order scattering cross section, the total sensitivity coefficient calculated by the independently developed SUFR program is in good agreement with the ERANOS and MCNP results. The maximum deviation does not exceed 3.22%. At the same time, the accuracy of the uncertainty analysis of the effective multiplication factor caused by the elastic scattering reaction channel of ²³⁸U and the inelastic scattering of ⁵⁶Fe has also been greatly improved. Therefore, the calculation of fast reactor sensitivity coefficient needs to consider the influence of high-order scattering cross section. At the same time, the sensitivity and uncertainty analysis program SUFR is developed correctly. The technical route for the sensitivity coefficient of fast energy spectrum reactors is feasible, and the calculation accuracy is the same as that of the famous international program.

Study on Irradiation Effect for Value of Ag-In-Cd Control Rods

Zhang Lidong, Zhao Jun

2021, 42(3): 55-59. doi: 10.13832/j.jnpe.2021.03.0055

Abstract(253) PDF(74)

Abstract:
In order to study the main neutron absorber nuclides of the Ag-In-Cd control rods in the reactor and its effect on the control rod value, Monte Carlo method is used to simulate the burnup of the main nuclides in Ag-In-Cd control rods during the reactor operation, and the neutron flux in the control rod and the macroscopic cross-section are analyzed, to investigate the value of the control rods before and after irradiation. The results show that the nuclide ¹¹³Cd in control rods decreases sharply with the increasing of the irradiation time, while ¹⁰⁷Ag, ¹⁰⁹Ag and ¹¹⁵In decrease slowly; the macroscopic cross-section of the control rods decreases, while the macroscopic cross-section of ¹⁰⁷Ag, ¹⁰⁹Ag and ¹¹⁵In increase; the neutron flux in the control rods increases, hence the neutron absorber rates in the Ag-In-Cd equal to that in the no-irradiation control rods.

Research of Water Packing Numerical Problem in Thermal Hydraulic System Code

Guo Yingran, Li Jiangkuan, Lin Meng, Yang Yanhua, Huang Tao

2021, 42(3): 59-63. doi: 10.13832/j.jnpe.2021.03.0059

Abstract(470) PDF(102)

Abstract:
Because of the discontinuous change in the compressibility between a two-phase mixture with a small void fraction and the pure liquid phase and the discrete method of the discrete momentum equation, the thermal hydraulic system program based on the two-fluid six-equation method may calculate a fictitious pressure peak when the steam is about to disappear from the volume or the water is about to fill the volume. In this paper, the thermal hydraulic system analysis program RELAP5 is used as a reference to analyze and mitigate the water packing problem. The detailed detection logic and mitigation scheme are given, and applied in the calculation and analysis of the Pryor’s Pipe Problem and the Condensation Test. The results show that the implementation of the water packing mitigation scheme can alleviate the pressure transient effects caused by the numerical method in the two-fluid program, which can significantly reduce the pressure peaks and avoid the emergence of severely distorted transient solutions. The water packing mitigation scheme slows down the peak of this fictitious pressure, which is helpful to improve the stability of program calculation. In view of this problem, this method can provide a reference for the development of the same type of system program and model optimization.

Strategies in Station Blackout Accident for Small Modular Reactors

Qiu Zhifang, Li Feng, Deng Jian, Cheng Kun, Du Zhengyu, Wu Lingyan

2021, 42(3): 64-69. doi: 10.13832/j.jnpe.2021.03.0064

Abstract(223) PDF(43)

Abstract:
The ability of nuclear power plants to deal with Station Blackout (SBO) accident has attracted much attention after Fukushima accident. Whether there is a sufficient ability to mitigate SBO accident has become crucial to measure the safety performance of a nuclear power plant. As a new type of reactor, small modular reactor requires higher safety performance. The mitigating effect of passive residual heat removal system and passive core cooling system for SBO accident of ACP100 is studied in this paper. The results indicate that ACP100 is with multiple strategies to deal with SBO accident that rely little on reliable power supply. Long-term residual heat removal and resultant long-term coolability can be ensured for ACP100 by the residual heat removal system or passive core cooling system.

Analysis of High-Temperature Transients in Severe Accident Depressurization Valve of 100 MW Pressurized Water Reactors

Wang Xiaoji, Wu Lingjun, Wu Qing, Liu Lili, Peng Huanhuan, Zou Zhiqiang

2021, 42(3): 69-74.

Abstract(227) PDF(43)

Abstract:
Due to the severe conditions in severe accidents of the nuclear power plants, the depressurization valves may experience a high temperature transient that the valves cannot withstand and may fail during the depressurization process. In this paper, the typical severe accident sequences that have a certain envelope and contain various typical severe accident phenomena are selected from the accident sequences that may lead to high-pressure melt ejection accident. According to the accident sequence, the high-temperature transient calculation was carried out considering the valves opening time range in severe accident management, and the water level of the pressurizer at the time of valve opening was analyzed for important influencing factors. Finally, the working conditions of the 100 MW nuclear power plant severe accident depressurization valves with typical and certain envelope characteristics are determined, and the maximum possible temperature and the variation temperature curve of fluid passing through the valves before and after valve opening are provided. It provides an important basis for equipment identification and function application of depressurization valves in severe accidents.

Static Elastoplastic Model of Metal Matrix Dispersion Fuel Element under Unstable Swelling Condition

Chen Hongsheng, Long Chongsheng, Xiao Hongxing

2021, 42(3): 74-80. doi: 10.13832/j.jnpe.2021.03.0074

Abstract(324) PDF(65)

Abstract:
For the unstable swelling of metal matrix dispersion fuel element induced by the cracking of metal matrix, the static elastoplastic model of crack surface was established without considering viscoplastic deformation, and this model was verified by the finite element simulation. The primary deformation mode transformed from elastic deformation into plastic deformation after the metal matrix completely yielded. The elastic and plastic deformation models of crack surface were established respectively based on the primary deformation modes of metal matrix. According to the balance between internal stress and bending moment, the critical transformation condition of elastic/plastic deformation was obtained. The calculated results of elastic and plastic deformation models fitted well with the finite element simulation, verifying the availability of the static elastoplastic model of metal matrix dispersion fuel under unstable swelling condition.

Evaluation of Effect of Fuel Pellet Manufacturing Parameters on Fuel Rod Performance by Numerical Fitting Method

Wang Kun, Zhang Kun, Xing Shuo, He Liang, Yin Mingyang

2021, 42(3): 80-85. doi: 10.13832/j.jnpe.2021.03.0080

Abstract(248) PDF(84)

Abstract:
Based on the theoretical calculation model, the manufacturing parameters related to the performance of fuel rods are obtained. With the help of FUPAC fuel rod performance analysis software, the sensitivity analysis of the parameters is carried out one by one, and the key parameters affecting the performance of fuel rods are selected. Based on a large number of calculation data of sensitivity analysis, the relationship function between the key parameters and the fuel rod performance is obtained by numerical fitting method, and the rapid and accurate evaluation of the effect of the manufacturing parameters of the pellet on the fuel rod performance is realized. The conclusion is that the numerical fitting method can effectively analyze the effect of fuel pellet manufacturing parameters on fuel rod performance through comparing the results of numerical fitting method and professional software analysis.

Development of Ultrasonic Transducer for Rod Cluster Control Assembly

Jin Xiaoming, Sun Jiawei, Li Bingqian, Hu Chenxu

2021, 42(3): 85-89. doi: 10.13832/j.jnpe.2021.03.0085

Abstract(310) PDF(61)

Abstract:
It is necessary to detect the defects of Rod Cluster Control Assembly used in nuclear power plant reactors to ensure the successful implementation in service inspection and reduce the cost of nondestructive testing. The ultrasonic probe for Rod Cluster Control Assembly (RCCA) inspection is developed independently. This paper describes the development process of ultrasonic probe about 15MHz-Φ4mm-H4mm-OD6mm. The development process is described in detail in three aspects: piezoelectric wafer, acoustic les and backing. The performance of the ultrasonic probe was tested. The number of pulse cycles is 1.5 weeks and the frequency band width is 105%. The defect test results are clearly visible, and the sensitivity and signal-to-noise ratio meet the inspection requirements in the simulation and inspection of the ultrasonic probe. It can completely replace the imported products with domestic ones.

A Hybrid CFD and Quasi-Static Theory Method for Fluidelastic Instability Prediction of a Tube Bundle

Song Lekun, Zhao Xielin, Zhou Jinxiong, Ye Xianhui, Feng Zhipeng, Xiong Furui

2021, 42(3): 89-94. doi: 10.13832/j.jnpe.2021.03.0090

Abstract(277) PDF(58)

Abstract:
In order to develop a method for predicting the fluidelastic instability of the tube bundle without relying on experiments, a hybrid fluidelastic instability prediction method for tube bundle is proposed by using CFD to capture drag and lift coefficients and their spatial derivatives, and substituting them into the quasi-static theory for fluidelastic instability. The developed method accounts for the instability of a tube bundle for both cross-flow and in-flow directions. Taking the normal triangle tube bundle as a typical example, the fluidelastic instability of two pitch ratios was performed. The results show that the prediction method of the elastic tube bundle fluidelastic instability based on the CFD and quasi-static theory hybrid can obtain the critical velocity of fluidelastic instability of the tube bundle without experimentation. The fluidelastic instability results calculated by this method are in good agreement with experimental results in literature.

Comparative Study on Fatigue Crack Growth Performance of Nuclear Stainless Steel Weld Joints and Base Metal

Chang Haijun

2021, 42(3): 96-103. doi: 10.13832/j.jnpe.2021.03.0096

Abstract(247) PDF(38)

Abstract:
Welded joints are widely used on the pipe sockets in nuclear power plants, and fatigue cracks are one of the important causes resulting in the failure of welded joints. Therefore, it is of great significance to study the fatigue crack propagation and life prediction methods for welding zone materials to accurately predict the life of welded joints. This paper takes the commonly used 304L stainless steel weld material in nuclear power plants as the object to study the effects of different load ratios and different sampling directions on the fatigue crack growth rate. Based on the test data, the fatigue crack growth rate model of the weld material is established, and it is compared with the austenite steel in ASME standard. The results show that different sampling directions have little effect on the fatigue crack growth rate of the weld, but the load ratio has a greater impact on it. At a lower load ratio, the fatigue crack growth rate of the weld is higher than that of the base metal before a certain ?K value, and thereafter it is lower than that of the base metal thereafter, but vise versa with higher load ratios.

Research and Application of 3D Visual Reinforcement System for Nuclear Power Containment Shell

Zhang Jie, Zhou Jianqiu, Xu Xinwei, Liu Quanchang

2021, 42(3): 103-108. doi: 10.13832/j.jnpe.2021.03.0103

Abstract(127) PDF(44)

Abstract:
Based on the nuclear power project under construction, a three-dimensional visualized reinforcement system for nuclear power containment is designed and developed on the basis of PDMS which is an international general plant design software, and a digitized reinforcement algorithm is studied. The visualization, digitalization and automation of reinforcement design of containment shell are carried out by program-driven reinforcement calculation equation and based on 3D data to realize the highly efficient output of engineering data such as two-dimensional drawing. The application of the research results in the project not only improves the accuracy and efficiency of the three-dimensional design, but also greatly improves the design quality, which has a certain significance of popularization for application.

Study on Countermeasures for Effect of Flow-Induced Vibration Test of Reactor Internals in First Demonstration Project of HPR1000 on Total Construction Period

Wang Qilong, Ma Ying, Xing Hui, Sun Chuanyi

2021, 42(3): 108-116. doi: 10.13832/j.jnpe.2021.03.0108

Abstract(194) PDF(51)

Abstract:
As the FOAK, HPR1000 is required to carry out the in-reactor measurement of flow induced vibration of reactor internals based on the regulation Comprehensive Vibration Assessment Program for Reactor Internals during Pre-operational and Start-up Testing (RG1.20) . The evaluation shows that this will result in additional 2.7 mouths in the construction period in the critical path. In order to eliminate the effect of test-related work on the construction period, this paper proposes a feasible optimization scheme by analyzing the influencing factors, countermeasures, risks and the associated effects after taking these measures. The analysis results indicate that the effect of test-related work on the construction period (62 months) of FOAK project can be eliminated by adjusting the logic of critical path and duration optimization.

Study on Small Power Improving of HPR1000

Xiang Meiqiong, Zhu Jialiang, Liu Yanyang, Qing Xianguo, He Zhengxi, Wu Qian, Zhu Biwei, Lyu Xin

2021, 42(3): 115-122. doi: 10.13832/j.jnpe.2021.03.0115

Abstract(347) PDF(81)

Abstract:
The thermal power accuracy of HPR1000 was analyzed, and the contribution of steam generator outlet pressure measurement accuracy, feed water temperature measurement accuracy and feed water flow measurement accuracy to HPR1000 thermal power accuracy was calculated. The quantitative data proves that the main feedwater flow measurement accuracy has the greatest influence on the thermal power calculation accuracy. Considering the current problem of low accuracy and the deterioration of the orifice flowmeter due to long-term use, a high-precision (0.3%) ultrasonic flowmeter is proposed to measure the main feed water flow. The calculation shows that the ultrasonic flowmeter can achieve a power increase of 0.97%.

Research on Double Containment Annulus Leakage Rate Test

He Rui, Shen Dongming, Li Shaochun, Chen Wei, Huang Xiaoming

2021, 42(3): 121-126. doi: 10.13832/j.jnpe.2021.03.0121

Abstract(279) PDF(41)

Abstract:
The tightness of the double containment annulus is critical to the safety of nuclear power plants, and the test procedure and data analysis should be reliable and reasonable. The formulation of the pressure in double containment annulus versus time is deduced in this paper with the foundation of the quadratic function law of pressure difference and the rate of air flow, which is used to calculate the leakage rate of the annulus at specified time. The fitting of annulus leakage rate against pressure difference is resulted from the analysis of the property of reference point for the measurement of annulus negative pressure. The quadratic theory is verified with project measured data, and is optimised with a corrected pressure difference, which performs better, and is recommended in the calculation of annulus leakage rate test.

Application of SPAR-H Method in Human Reliability Analysis of Digital Nuclear Power Plants

Qing Tao, , Liu Zhaopeng, Zhang Li, Tang Yaqin, Hu Hong, Zang Jing

2021, 42(3): 126-132. doi: 10.13832/j.jnpe.2021.03.0126

Abstract(508) PDF(69)

Abstract:
The applicability of SPAR-H method in the digital nuclear power plant has not been fully studied. This paper studied the operator behavior characteristics of the digital nuclear power plant and the application of SPAR-H method in Lingdong Nuclear Power Plant. The study results show that SPAR-H method has some deficiencies, such as over-conservative analysis results, incomplete cognitive process, and over-sensitivity of partial PSF when applied to digital nuclear power plants. In view of the above shortcomings, some suggestions are put forward for SPAR-H method, such as defining the PSF criterion, perfecting the cognitive model of SPAR-H method, and establishing the human factors database, and thus SPAR-H method can be more suitable for HRA of digital nuclear power plants.

Research on Covert Attack Method in Large Pressurized Heavy Water Reactors

Zhang Yan, Fan Dengning, Huang Yu, Wang Dongfeng, Xu Peihao

2021, 42(3): 132-140. doi: 10.13832/j.jnpe.2021.03.0132

Abstract(407) PDF(39)

Abstract:
In order to facilitate the research and development of the security defense system for large pressurized heavy water reactors (PHWR), this paper studies the potential attack mode in PHWR networked control system and proposes a covert attack approach based on Gaussian process regression model optimized by salp swarm algorithm. In this method, when the false data is injected into the PHWR networked control system, the system identification is addressed by optimizing the Gaussian process regression algorithm and a high-precision estimation model of the attacked area in PHWR is obtained, and then the estimation model is used to realize the covert attack. The simulation results show that the attack method not only causes some damage to PHWR, but also has high concealment performance.

Research on Feedforward Compensation for Steam Generator Level Control System Manual/Automatic Switch

Xu Ying, Chen Jiancai, Yu Hang, Wang Zhixian

2021, 42(3): 140-145. doi: 10.13832/j.jnpe.2021.03.0140

Abstract(243) PDF(54)

Abstract:
Based on the current design of the replication loop for the steam generator level control system, after the main feed water flow regulating valve is switched from manual mode to automatic mode, the calculation basis of the level controller is the steam water mismatch signal during switching, which causes the control system to lose the feed-forward function of fast regulating feed water flow. Combined with the HH reactor trip event in the steam generator level of unit 4 in Yangjiang Nuclear Power Plant, two optimization schemes are proposed for the manual/automatic switching operation scheme and systematic design. For the optimization of the operation mode,, before the main feed water flow regulating valve is put into operation automatically, the steam water flow is manually balanced. For the optimization of the systematic design, the steps for the steam water mismatch judgment and the feed-forward compensation are added. Through the liquid level disturbance test of the steam generator in unit 4 of Yangjiang Nuclear Power Plant, it is proved that the optimization scheme can effectively improve the regulating speed of the control system and reduce the overshoot, which has an important contribution to the safe operation level of the unit.

Research of Quantification Method of Risk in Seismic Probabilistic Safety Analysis in Nuclear Power Plants

Jing Xu, Xiao Jun

2021, 42(3): 145-150. doi: 10.13832/j.jnpe.2021.03.0145

Abstract(274) PDF(45)

Abstract:
The current status of quantification method and tools for seismic probabilistic safety assessment (PSA) in nuclear power plants was discussed, and the challenges faced by quantitative tools and the issues need to be resolved was suggested. A quantitative method based on the nature of probability theory was proposed. The application process of the calculation method was demonstrated, taking the results of multi-plan probabilistic seismic hazard analysis (PSHA) and the minimum cut set given by the seismic response analysis of a nuclear power plant in China as inputs, and then the effect of ground motion parameters and confidence parameters on the quantitative results was analyzed. The results demonstrate that the Latin hypercube sampling for the confidence parameter can give a stable estimate of the core damage frequency caused by earthquake (SCDF) in the nuclear power plant even the number of samples is small; in general, the equipment failure contributes the most to the SCDF; the impact of structure failure is relatively small; the contribution of the annual occurrence frequency of ground motion parameter to SCDF needs to be specifically analyzed according to the location of the project site.

Study on Level 2 PSA Release Categories and Selection of Representative Accident Sequence in Nuclear Power Plants

Zhang Jiajia, He Dongyu, Gong Yu, Luo Yong, Chen Peng, Chen Yingying

2021, 42(3): 149-154. doi: 10.13832/j.jnpe.2021.03.0150

Abstract(210) PDF(42)

Abstract:
Domestic nuclear power projects such as AP1000, EPR and HPR1000 have used the level 2 PSA source terms for the emergency input, but there is no clear operational methods for the division of level 2 PSA release categories and the selection of representative accident sequences for each release category, and further research is required. Comparing the level 2 PSA release categories and the selection of representative accident sequences for each release category of domestic advanced nuclear power plants, taking a domestic generation = 3 * ROMAN * MERGEFORMAT III advanced pressurized water reactor nuclear power plant as an example, the source terms were calculated for 4 different accident sequences that selected from a release category based on frequency and consequences. The results show that the source results of different accident sequences are quite different. It is recommended that the release category be divided into application-oriented and iterative according to the analysis purpose, and multiple accident sequences should be selected for the same release category for comparative analysis to demonstrate reasonable release categories and representativeness of accident sequences.

Model Study on Hydrogen Stratification Behavior within a Containment

Peng Cheng, Deng Jian

2021, 42(3): 155-160. doi: 10.13832/j.jnpe.2021.03.0155

Abstract(562) PDF(53)

Abstract:
In this paper, with theoretical modeling and experimental correlation, a semi-empirical model which can predict the hydrogen distribution characteristics has been proposed, based on the dominant factors of interaction among the inertial force, viscous force and buoyancy under the injection of both steam and hydrogen. The rationality of the model is verified by comparing with the experimental data under the injection of medium and high content of steam. All the theoretical work can provide extra support to further develop the hydrogen distribution model coupled with condensation effect within the containment in the future. Moreover, its application in the study of the typical hydrogen behavior on small-scaled containment model of CAP1400 demonstrates that there may be reservoir of light gas, gradient layers and stagnation region along the vertical direction, which is in line with the findings by international benchmark tests (ISP47).

Study on Online Monitoring of Equipment Condition Based on Local Outlier Factor and Artificial Neural Networks Model

Shen Jiangfei, Li Huaizhou, Huang Lijun, Mao Xiaoming, Zhang Sheng

2021, 42(3): 160-166. doi: 10.13832/j.jnpe.2021.03.0160

Abstract(370) PDF(68)

Abstract:
The centralized online monitoring technology plays the most important role in nuclear power plants for the safety of major equipments and economic operation. In order to solve the false alarm and alarm failure problems in the traditional online monitoring, a new artificial intelligence monitoring method based on the local outlier factor and artificial neural networks model is put forward in this paper. This method is one of the multiple parameter dynamic threshold detection method. Firstly, a group of monitoring parameters of equipment is selected by analyzing the failure modes and failure phenomena of equipment. Secondly, enough data of this group of parameters needs to be collected and the abnormal data needs to be screened out. Thirdly, all the selected data is used to calculate the local outlier factor, and then the neural network model will be established by inputting the selected data and the local outlier factor. Finally, the neural network model can be used to assess the equipment health index with the real-time data of equipment parameters as input, and the health index represents the real-time health of equipment. In this paper, this method is used to develop a monitoring model of circulating water pump. In order to verify the validity of the model, enough monitoring data of healthy equipment and malfunction equipment are used to verify the monitoring results. The results show that the method can provide a pre-alarm for the early failure of the equipment with low false alarm rate, greatly improving the monitoring efficiency.

Disposal of Abnormal Tension Event of Main Pump Snubber Tie Rod

Chen Sun, Zhang Yihan, Zhao Xiaohong, Li Shilei

2021, 42(3): 166-171. doi: 10.13832/j.jnpe.2021.03.0166

Abstract(199) PDF(34)

Abstract:
Abnormal tension occurs on the main pump snubber tie rod of a nuclear station when it is loaded by bolt stretcher. In order to solve this problem, this paper uses the methods of cause elimination and analytical calculation to diagnoze the fault. It is identified that the main reason for abnormal tension is the short thread length. In view of this problem, the scheme of adjusting the screw thread length is proposed and verified by the finite element method, and the stress analysis of the anchor rod thread also meets the requirements of the code. The handling of this event can guarantee the construction progress on site and avoid great economic loss.

Preliminary Study of General Design of Floating Nuclear Power Plants

Chen Yanxia, Zhu Chenghua, Guo Jian, You Xiaojian, Zhang Jincai, Tan Mei, Li Pengfan

2021, 42(3): 171-177. doi: 10.13832/j.jnpe.2021.03.0171

Abstract(386) PDF(83)

Abstract:
Based on the analysis of the advantages and disadvantages of the existing reactor types in the world and the application on ships, this paper suggests that the mature PWR can be used in the floating nuclear power plants on the sea, and the principle suggestions on the reactor power and refueling cycle are given. Taking the ship type floating nuclear power plant of single point mooring type as an example, the principle of compartment division is expounded, and the division is carried out according to the main functions of each cabin. The main restriction factors of the main scale are analyzed, and the general layout principle is expounded. The basic design principles of containment, safety enclosure, radioactive waste management system and biological shielding in the reactor compartment are emphatically introduced. At the same time, the design principles of some key systems such as secondary circuit, control room, power system and physical protection are introduced.

Study on Design and Safety Analysis of Spent Fuel Storage Grid for Marine Nuclear Power Platform

Mei Zhen, Sun Fujiang, Zhu Gang, Yu Ying, Chen Juan, Lu You

2021, 42(3): 177-183. doi: 10.13832/j.jnpe.2021.03.0177

Abstract(122) PDF(35)

Abstract:
For the problem in the safety assurance for long-term maritime storage of spent fuel for marine nuclear power platform, this paper improved the fixed form between the fuel assembly and the storage cell, optimized the connection form between the storage cell and the spent fuel storage grid frame body, and added a buffer structure between the spent fuel storage grid and the wall of the spent fuel pool. In this way, a spent fuel storage grid meeting the design criteria and adapting to the marine environment is designed. Monte Carlo code MCNP-5, computational fluid dynamics (CFD) software Fluent 14.0 and finite element analysis software ANSYS 17.0 are used to simulate the criticality, the thermal hydraulics and the structural of this storage grid. The results show that the design of the storage grid is reasonable and safe. Therefore, the spent fuel storage grid designed in this study can provide a solution for the maritime storage of spent fuel in floating nuclear power plants.

Fault Diagnosis of Vibration Induced by Fluid of 100D Main Pump for CPR1000 Unit

Shu Xiangting, Yang Zhang, Xu Yizhe, Jiang Yanlong

2021, 42(3): 183-188. doi: 10.13832/j.jnpe.2021.03.0183

Abstract(214) PDF(60)

Abstract:
The vibration phenomenon of 100D main pump under various operation modes for CPR1000 unit show that, when the main pump motor is in normal shutdown condition of steam generator cooling or residual heat removal cooling, the amplitude value of the main pump motor tile often has a large range of impact fluctuation or even triggers high vibration alarm. According to the principle of mechanical vibration, the frequency-domain and time-domain characteristics of motor bearing vibration and main pump shaft displacement signals are comprehensively analyzed. The vibration amplitude fluctuation is affected by the low-frequency random vibration of about 7~9 Hz. The signal collected by the vibration and noise monitoring system is used to analyze the low frequency vibration corresponding to the core basket beam vibration induced by the main coolant flow in the primary circuit. According to the theory of fluid induced vibration, the main factors affecting the vibration fluctuation of the main pump motor are analyzed and verified by the historical operation records of the main pump. The suggestion of optimizing the operation strategy of CPR1000 unit to alleviate the vibration fluctuation of main pump motor is put forward systematically, which provides a reference for the safe and stable operation of the main pump.

Comprehensive Leakage Diagnosis Technology of Primary Loop Pressure Boundary of Nuclear Power Plants Based on Leakage Monitoring Data Synthesis

Ling Jun, Yang Yutao, Li Hongxia, Zang Yiming, Tan Ke, Yuan Jingqi

2021, 42(3): 188-193. doi: 10.13832/j.jnpe.2021.03.0188

Abstract(373) PDF(73)

Abstract:
The leakage monitoring system is used to monitor the integrity of the reactor coolant system pressure boundary (RCPB), and is also a prerequisite for the application of leak before break (LBB) technology. Leakage comprehensive diagnosis is the core function of leakage monitoring system. In this paper, the technical scheme of leakage comprehensive diagnosis is constructed from six aspects: system availability, data reliability, single meter leakage alarm, leakage comprehensive diagnosis, alarm response strategy and automatic calculation of leakage rate periodic test. The sensitivity and accuracy of leakage monitoring system are important performance indicators of the comprehensive leakage diagnosis technology, and also the key requirement of the integrity of RCPB and the application of LBB technology. Firstly, the conservative threshold range of triggering single instrument alarm is determined to ensure the sensitivity of detection. Then, the effective single instrument alarm threshold is checked and adjusted by the leakage comprehensive diagnosis technical scheme to ensure the accuracy of the alarm. Through theoretical calculation, data analysis and multi signal consistency judgment, the leakage monitoring system can timely and accurately diagnose the leakage. The intelligent of leakage diagnosis technology is fully applied to reduce the workload of review and alarm of operators.

Research on Efficient Compound Decontamination Technology for Decommissioning of Highly Radioactive Hot Cell

Jia Haopeng, Teng Lei, , Wang Shuai

2021, 42(3): 193-197. doi: 10.13832/j.jnpe.2021.03.0193

Abstract(246) PDF(50)

Abstract:
The highly radioactive hot cell is used as an auxiliary facility for the irradiation inspection of reactor materials, and the hot cells are with high radiation level, complicated structure and great difficulty in decontamination. In view of the particularity and complexity of the decontamination of the decommissioned stainless steel shell in the strong heat release chamber, this paper is carrying out three single-item decontamination tests and decontamination process test studies on the high-pressure water jet decontamination, the strippable film decontamination and the mechanical polishing decontamination. On the basis of this, an innovative composite decontamination process for stainless steel shells in a strong heat release chamber was innovatively proposed. The engineering decontamination practice verified that, the average level of contamination on the surface of the stainless steel cladding after the decontamination is below 40 Bq/cm². The factor is as high as 110 or more, reaching the domestic advanced level. The research and development of the high-efficiency compound decontamination technology in the hot cell solves the technical problem of the decontamination of the stainless steel shell surface in the highly radioactive hot cell, reduces the exposure dose of workers in the decommissioning stage, protects the safety of workers and the environment, and has significant economic and social benefit.

Research on Accident Diagnosis Method for Reactor Primary Circuit System Based on SDG and PCA

Ma Jie, Zhang Longfei, Yu Ren, Peng Qiao, Hu Pengfei

2021, 42(3): 197-203. doi: 10.13832/j.jnpe.2021.03.0197

Abstract(270) PDF(59)

Abstract:

Research on Hot Leg Temperature Mixing and Measurement Characteristics of Generation III Advanced Reactor

Ren Chunming, Du Sijia, Deng Jian, Wu Qing, Xin Sufang, Hu Ying, Liu Xiaobo

2021, 42(3): 203-207. doi: 10.13832/j.jnpe.2021.03.0203

Abstract(205) PDF(39)

Abstract:
In order to predict the rationality of thermometers setting on hot leg of Generation Ⅲ Advanced Reactor,the hot leg temperature mixing and measurement characteristics was studied for conditions with different core outlet temperature and flowrate distributions, using computational fluid dynamic (CFD) analysis technology, with model consist of volume from core outlet to location of thermometers on hot legs. The results show that obvious temperature stratification of the coolant in hot legs exists, while the deviation of the average temperature measured by the thermometer is relatively small compared with the average temperature of the coolant section. Therefore, the thermometer setting on hot legs for Generation Ⅲ Advanced Reactor is reasonable and the temperature of coolant can be effectively measured.

First Collision Compensation Technique for Leakage Source

Tang Xiao, Li Qing, Chen Zhang, Chai Xiaoming, Tu Xiaolan, Wang Liangzi, Li Mancang

2021, 42(3): 207-211. doi: 10.13832/j.jnpe.2021.03.0207

Abstract(144) PDF(39)

Abstract:
In order to solve the problem of convergence instability of two-dimensional / one-dimensional neutron transport calculation, the first collision compensation for leakage source technique is proposed. The source term is equivalent to the scattering source of each region by the first collision method, which is equivalent to distributing the local isolated source into the whole wide space, so as to reduce the influence of the radiation effect of the leakage source. The single energy correction method is applied to simplify the calculation, and to improve the convergence, stability and accuracy of the two-dimensional / one-dimensional neutron transport calculation.

Review on Development of Critical Heat Flux Mechanistic Model

Liu Wei, Peng Shinian, Jiang Guangming, Liu Yu, Deng Jian, Hu Ying, Liu Xiaobo

2021, 42(3): 211-218. doi: 10.13832/j.jnpe.2021.03.0211

Abstract(1769) PDF(222)

Abstract:
In order to clarify the development of critical heat flux (CHF) mechanistic model and promote the CHF experimental and theoretical research, the existing achievements and progress of CHF mechanistic model are systematically reviewed and the basic assumptions and detailed modeling process of each model are analyzed in this study. In addition, the existing problems and corresponding possible solutions of each model are demonstrated. This study can provide a basis for current understanding and experimental approaches of CHF phenomenon.

Development and Application of COSINE Reactor Monte Carlo Code cosRMC

Yu Hui, Quan Guoping, Qin Yao, Yan Yiman, Chen Yixue

2021, 42(3): 218-224. doi: 10.13832/j.jnpe.2021.03.0218

Abstract(323) PDF(57)

Abstract:
The self-developed COSINE Monte Carlo code cosRMC is designed for the analysis of the reactor core and the radiation shielding. The typical functions as transport simulation, burnup calculation, group constants generation, sensitivity & uncertainty analysis and visual modeling have been sufficiently researched. A lot of V&V work was carried out, and the recent application in AP1000 PWR and China Fusion Engineering Test Reactor (CFETR) is introduced in this paper to show the capability of cosRMC. The simulation results show the maximum absolute deviation is 89.9 ×10^-5 for eigenvalue and 2.1% for power distribution of the 21 types of assemblies and core models of AP1000, while the maximum absolute deviation is 0.6% of the CFETR thorium proliferation ratio. The typical work further exhibits that cosRMC software can meet the computational needs of the large and complex models such as PWR and Fusion Reactor with high accuracy, and the visual modeling tool imbedded can effectively improve the modeling efficiency.

Research on Optimization of Core Fuel Management for Units 1-4 of Tianwan NPP

Guo Zhipeng, Wu Jinying, Xu Min, Zhang Haoran, Yi Xuan, Huang Peng, Ye Liusuo

2021, 42(3): 224-229. doi: 10.13832/j.jnpe.2021.03.0224

Abstract(331) PDF(268)

Abstract:
This paper has done research on the optimization of core fuel management for units 1-4 of Tianwan NPP by KASKAD program package to unify the fuel management schemes. In the optimizing schemes, less varieties and quantities of fresh fuel assembles are used, however the average enrichment of new FAs is higher to increase the length of equilibrium cycle. All the safety parameters of the optimizing schemes according to the core calculation results satisfy the requirements of limiting values. The optimizing schemes are with good flexibility, and can increase the utilization of the fuel as well as increase the economic benefit of the plant, and have good engineering application value.

Generalized Perturbation-Theory-Based Sensitivity Analysis with CMFD Acceleration

Wu Qu, Peng Xingjie, Yu Yingrui, Li Qing

2021, 42(3): 229-233.

Abstract(259) PDF(38)

Abstract:
To perform the generalized sensitivity analysis for nuclear data in a reactor physics design code, KYLIN-Ⅱ, the generalized perturbation theory is adopted and several generalized fix-source equations with the orthogonal definite condition need to be solved when the sensitivity coefficients are figured out. Besides, the paper develops a new approach, CMFD-based generalized fix-source equation solution, to accelerate the convergence. The convergence efficiency of the generalized fixed-source equation is improved by roughly 4.3 times, and the sensitivity coefficients calculated by the GPT accord with those calculated by the direct perturbation theory, which demonstrates the sensitivity analysis ability in KYLIN-Ⅱ.

2021 Vol. 42, No. 3