Advance Search

2023 Vol. 44, No. 6

Reactor Physics
Conceptual Design Optimization of Uranium-zirconium Alloy Fuel Core for Modular Lead-based Fast Reactor
Lou Lei, Wang Lianjie, Chen Chang, Zhao Chen, Zhou Bingyan, Yan Mingyu, Ma Dangwei
2023, 44(6): 1-8. doi: 10.13832/j.jnpe.2023.06.0001
Abstract(151) HTML (28) PDF(76)
Abstract:
In order to deeply study the physical properties of lead-based fast reactor, one of the reactor types of the fourth generation nuclear energy system, and further improve the safety and economy of modular lead-based fast reactor, the core characteristics of modular lead-based fast reactor with different power levels loaded with uranium-zirconium alloy fuel were studied, and it was found that when the core power was raised to a certain level, the breeding advantage of the core could not be fully released within the specified life. Based on this phenomenon, the conceptual design of modular lead-based fast reactor uranium-zirconium alloy fuel core is optimized. Based on the core power level and service life, the appropriate ratio of grid spacing to rod diameter and the effective density of fuel core are selected, and the breeding performance of core is adjusted by adjusting the uranium loading per unit volume and the 235U loading. Finally, the core reactivity changes are basically matched with the core power and service life, so that the core reactivity hardly changes during the service life. The optimization not only reduces the difficulty of core reactivity control, but also makes full use of the breeding performance of the core. At the same time, the reasonable ratio of grid spacing to rod diameter provides safety and design margin for thermal analysis of the core, which effectively improves the economy and safety of the core.
Anisotropic SP3 Two-Step Method for Reactor Cores with Strong Absorbers
Li Yunzhao, Qin Junwei, Xia Fan, Wu Hongchun, Cao Liangzhi
2023, 44(6): 9-15. doi: 10.13832/j.jnpe.2023.06.0009
Abstract(47) HTML (10) PDF(21)
Abstract:
The strong absorber in the core of nuclear reactor, especially the movable control rod, will significantly enhance the angular anisotropy effect of neutron angular fluence in the core. The traditional isotropic SP3 two-step method cannot capture the feature effectively, therefore, the anisotropic SP3 two-step method needs to be constructed. Firstly, this paper studies the anisotropic effect from two aspects; the anisotropic SP3 equation is derived, and a homogenization model suitable for anisotropic SP3 equation is established. The methodology proposed in this paper was validated and analyzed through the Material Test Reactor (MTR). The results show that the anisotropy SP3 equation and its homogenization method in this paper have smaller deviation than the traditional calculation method, and the effective multiplication factor (keff) and power distribution are obviously improved. Therefore, the anisotropic SP3 two-step method in this paper can effectively deal with the core problem with strong absorbers.
Prediction of Critical Parameters of Reprocessing Non-uniform Conditions Based on Improved BP Neural-network
Sun Mingze, Cheng Yuting, Ma Xiaodi, Xia Zhaodong, Zhou Qi, Zhu Qingfu, Xue Xiaogang
2023, 44(6): 16-22. doi: 10.13832/j.jnpe.2023.06.0016
Abstract(65) HTML (11) PDF(16)
Abstract:
As the key process equipment of reprocessing plant, the extraction column and storage tank often have the condition of fluctuating solution concentration (i.e., non-uniform condition). When conducting critical safety analysis, technicians adopt the conservative method of enlarging concentration several times. Although this meets the conservative requirements, it introduces too much critical margin, which limits the treatment efficiency and capacity of reprocessing. In order to solve the above problems, based on the improved BP neural network method and the large-scale MC code MCNP, this study completed the gradient modeling of random concentration distribution for typical equipment structure size, and realized the critical safety analysis method of predicting effective proliferation factor (keff) based on concentration distribution. Data test results show that the average error of keff calculated under non-uniform conditions by this method is 1.82×10−4, and the convergence value of loss function MSE is 3.34×10−6, which is far smaller than the unimproved model (2.4450×10−4). At the same time, in comparison with the conservative method, the critical margin introduced by the proposed method is –1.31×10−3, which is much smaller than that of the traditional method (0.32951). The above results prove that the method in this study is more accurate and effective under the precondition of conservativeness, and provide a method reference for the critical safety analysis of reprocessing.
Treatment Method of Control Rod Cusping Effect in 2D/1D Coupled Characteristics Transport Calculation
Zhang Sifan, Liu Zhouyu, Zhao Hanbing, Chen Xing, Wang Bo
2023, 44(6): 23-31. doi: 10.13832/j.jnpe.2023.06.0023
Abstract(36) HTML (10) PDF(7)
Abstract:
On the basis of 2D/1D coupled characteristics transport, the treatment method of cusping effect of control rod movement are studied, including adaptive mesh method and bidirectional homogenization method. In the bidirectional homogenization method, the homogenization error is reduced by establishing radial flux profile and axial flux profile. The control rod movement function is implemented in high fidelity numerical reactor code NECP-X. The treatment of the control rod cusping effect is verified by the C5G7-TD benchmark and VERA #5 benchmark. The numerical results show that the adaptive mesh method and the bidirectional homogenization method both have high accuracy in the calculation of control rod value and core power. The control rod cusping effect caused by bidirectional homogenization can be well corrected by flux profile.
Three-Dimensional Pin-by-pin Transient Analysis for PWR-Core
Qin Junwei, Li Yunzhao, Wang Kunpeng, Wu Hongchun, Cao Liangzhi
2023, 44(6): 32-38. doi: 10.13832/j.jnpe.2023.06.0032
Abstract(65) HTML (6) PDF(23)
Abstract:
In order to ensure the safety of nuclear reactor in transient process, it is necessary to carry out three-dimensional transient analysis of the whole core to obtain the power distribution of fuel rods concerned in the safety analysis. In this paper, a theoretical model of the pin-by-pin neutronic-thermohydraulic coupling transient analysis for PWR-core is established: the three-dimensional pin-by-pin spatio-temporal neutron dynamics calculation of PWR-core is realized by fully implicit backward difference method and exponential function expansion nodal method; the multi-channel model of cell scale is used to calculate the thermal feedback of three-dimensional full core pin-by-pin; the Picard iteration is used to realize the iterative calculation of neutronic-thermohydraulic tight coupling; using the parallel technology of message passing interface (MPI) and the unified spatial domain decomposition method, the parallel calculation of neutronic-thermohydraulic coupling transient process is realized. Based on the theoretical model, the corresponding code called Bamboo-Transient 2.0 was developed and verified by the benchmark and multi-assemblies questions, and it was then applied to the transient analysis for commercial PWRs. The results show that the pin-by-pin transient analysis code has more fine results than that of the traditional coarse-mesh diffusion code based on assembly homogenization, and it can reduce the deviation of power distribution of pin-cell. Meanwhile, it can provide the distribution of state parameters at cell scale, which can directly meet the requirements of safety analysis.
Experimental Study on Leakage of High Temperature Sodium Heat Pipe in the Air
Liu Shuai, Zhou Yuan, Kang Mingming, Yuan Yuan, Du Zhengyu, He Xiaoqiang, Hu Wei
2023, 44(6): 39-44. doi: 10.13832/j.jnpe.2023.06.0039
Abstract(28) HTML (9) PDF(11)
Abstract:
After the leakage of high temperature sodium metal in the heat pipe cooled reactor, it will react with air to produce combustion and even explosion, which will endanger the safety of the reactor core. The research on sodium leakage of heat pipe at high temperature is faced with some problems, such as few experiments and unknown phenomena. In this paper, simulation experiments of sodium heat pipe top leakage was carried out. 15 g of sodium was put into a stainless steel pipe and heated to 904.8℃. The degree of danger was judged by measuring the pressure and temperature changes during the experiment. The results showed that the leakage of the heat pipe produced yellow flame and explosion sound. 0.081 MPa pressure pulse was detected in the reaction chamber, and the maximum temperature rise of the reaction chamber was 192.1℃. The experimental results include experimental phenomena and leakage process analysis, which provides a basis for the subsequent design and safety assessment of heat pipe cooling reactor.
Study of KRUSTY Thermal Expansion Negative Feedback Calculation Based on Unstructured-Mesh MCNP
Wang Lipeng, Cao Lu, Chen Sen, Zhang Xinyi, Jiang Duoyu, Hu Tianliang, Li Da, Chen Lixin, Jiang Xinbiao
2023, 44(6): 45-53. doi: 10.13832/j.jnpe.2023.06.0045
Abstract(59) HTML (13) PDF(28)
Abstract:
Thermal expansion negative feedback simulation of heat pipe reactor KRUSTY has always been a difficulty in calculation. Based on the unstructured mesh ablility of MCNP, the power distribution of KRUSTY unstructured mesh is directly input to ABAQUS, and the thermal mechanical coupling of KRUSTY is carried out by ABAQUS. The thermal deformation simulation, expansion reactivity feedback and density feedback of KRUSTY are studied under the unified unstructured mesh. The difference between non-uniform density and uniform density is studied. The results show that the thermal expansion effect brings more than 900pcm (pcm=10-5) negative feedback, and the special fuel deformation mainly occurs on the upper and outer edge surface, with a total displacement of about 0.9 cm. The total temperature difference of reactor core is small enough, only about 23 K. Moreover, the neutronics-thermo-mechanics coupling tends to make the core temperature distribution more uniform, and the reactor design can satisfy the single-point failure principle due to the redundant design of the heat pipe. Compared with the traditional CSG geometry, the unstructured-mesh Monte Carlo method can better simulate the thermal expansion effect of metal fuel reactor.
Thermal and Hydraulic
Numerical Study of the Influence of Rolling Motion on the Spacer Effect of Low Flow Convective Heat Transfer
Li Nan, Ding Guanqun, Xiao Yao, Li Junlong, Gu Hanyang
2023, 44(6): 54-62. doi: 10.13832/j.jnpe.2023.06.0054
Abstract(95) HTML (27) PDF(30)
Abstract:
A numerical study was carried out on the spacer effect of low flow convective heat transfer under rolling motion. The rolling model was verified based on the experimental data, then a model-based computational fluid dynamics (CFD) method was established. The results show that for the circumferential mean time-averaged heat transfer, the heat transfer downstream of the spacer is enhanced under low flow conditions. However, in the mixed convection deterioration recovery area and natural convection area, the spacer effect still has damping oscillation attenuation, and the oscillation amplitude decreases. For local time-averaged heat transfer, the hot spot always lies in the direction of vertical rolling axis, and the heat transfer is weakened in that direction under rolling motion, and the heat transfer weakening caused by the spacer in that direction is further enhanced. For the instantaneous heat transfer at the hot spot, the maximum weakening degree of the instantaneous heat transfer coefficient can reach 40% of the fullt developed steady state, which needs to be paid attention to in safety analysis.
Analysis of Multi-branch Natural Circulation Characteristics of Passive Residual Heat Removal System in NHR200-Ⅱ Reactor
Geng Yiwa, Liu Xiongbin, Li Xiaotian, Zhang Yajun
2023, 44(6): 63-70. doi: 10.13832/j.jnpe.2023.06.0063
Abstract(946) HTML (15) PDF(49)
Abstract:
There is uneven flow distribution in multi-branch natural circulation system. In order to further analyze the flow characteristics of the system, a simplified mathematical model of multi-branch parallel-channel natural circulation loop was established based on the passive residual heat removal system (PRHR) system of the NHR200-Ⅱ reactor test facility. The mechanism of the reverse flow phenomenon was analyzed by utilizing the pressure drop-flow rate diagram, and the effects of various factors on the number of reverse flow branches were discussed. The analysis results show that the thermal power of PRHR system is deteriorated as the reverse flow occurred, and increasing the elevation between the primary heat exchanger (PHE) and the air cooler (RHE) can suppress the reverse flow in the PHE branches. There is a critical elevation. When the elevation is less than the critical value, changing the resistance of PHE or RHE branches will not change the number of reverse flow PHE branches. When the elevation is greater than the critical value, increasing the resistance of PHE branches or decreasing the resistance of RHE branches can reduce the number of reverse flow branches until the reverse flow is completely suppressed.
Numerical Study on Flow and Heat Transfer Characteristics of Subcooled Boiling in 5×5 Petal-shaped Fuel Rod Assembly
Cai Weihua, Huang Zequan, Zhang Wenchao, Wei Zhisheng, Cui Jun, Jin Guangyuan
2023, 44(6): 71-79. doi: 10.13832/j.jnpe.2023.06.0071
Abstract(53) HTML (9) PDF(22)
Abstract:
Based on the Eulerian two-fluid model and the Rensselaer Polytechnic Institute (RPI) wall boiling model, and considering the fluid-solid coupling heat transfer in fuel rod assembly, the flow and heat transfer characteristics of subcooled boiling in the 5×5 petal-shaped fuel rod assembly under the condition of uniform volume heat source was studied, and the velocity field, temperature field, void fraction distribution and heat transfer coefficient distribution in different sub-channels were analyzed. The results show that the secondary flow intensity in the rod bundle channel changes periodically along the axial direction. Under subcooled boiling condition, the peak value of void fraction in petal-shape fuel assembly appears near the outlet. The bubbles are mainly generated at the elbow of the fuel rod and distribute eccentrically counterclockwise, and the volume fraction of vapor in the corner subchannel is obviously larger than that in the center subchannel. Under the simulated conditions in this paper, the maximum temperature of the pellet reaches 657.9 K. The area of high temperature zone of the fuel rod pellet increases gradually along the axial direction, and the coolant temperature in the corner subchannel is higher than that of the edge subchannel. The average coolant temperature of the central subchannel is the lowest, and the heat transfer coefficient of each subchannel fluctuates periodically along the axial direction.
Study on Particle Size Prediction of Sodium Atomization Based on Unstable Wave Theory
Xu Zhen, Hu Peizheng, Tong Lili, Cao Xuewu
2023, 44(6): 80-85. doi: 10.13832/j.jnpe.2023.06.0080
Abstract(37) HTML (14) PDF(15)
Abstract:
The particle size of sodium atomization is the key factor affecting the combustion intensity of sodium spray fire. Because of the unstable chemical properties of sodium, water is usually used to replace sodium to carry out atomization experiments, and it is necessary to obtain the particle size conversion relationship between them. In this paper, based on the unstable wave theory, the particle size similarity model of liquid atomization characterized by liquid physical parameters and atomization pressure difference is established. The applicability of the model is verified by different atomization experiments of various fluids from the perspectives of pressure difference and physical parameters. Furthermore, an experimental facility for water atomization using nozzle to simulate the crack is designed, the average particle size of water atomized droplets with pressure difference of 0.1~0.5 MPa is obtained, and the average particle size of sodium atomized droplets at 300~600℃ under different pressure differences is predicted. The particle size similarity model and the experimental device for sodium atomization simulation using water established in this paper can realize the prediction of sodium atomization particle size under different pressure differences, and provide reference for the research of Sodium atomization fire accidents.
Architecture Design of Two-Fluid Two-Pressure Thermal-Hydraulic System Analysis Code LOCUST 2.0
Xu Caihong, Yuan Hongsheng, Liu Huannan, Ju Zhongyun, Li Changying, Wang Ting, Li Jinggang
2023, 44(6): 86-94. doi: 10.13832/j.jnpe.2023.06.0086
Abstract(850) HTML (38) PDF(55)
Abstract:
China General Nuclear Power Corporation (CGN) is developing a new thermal hydraulic system analysis code LOCUST 2.0, which uses two-fluid two-field two-pressure seven-equation models, theoretically ensuring the strict well-posedness of conservation equations. The design application of LOCUST 2.0 is the safety analysis of loss-of-coolant accidents (LOCAs) for HPR1000, and its source codes have been preliminarily completed at present. In this paper, the conservation equations, numerical methods, software functions, and code architecture design are briefed introduced. Meanwhile, the calculation results of six typical test cases are given, and the results show that the code reasonably predicts the thermal-hydraulic processes and has good performance.
Study on Circumferential Non-Uniformity of Annular Fuel Outer Temperature Distribution in Lead-bismuth Cooled Tight Cell
Zeng Fulin, Zhang Xiaolong, Zhao Pengcheng
2023, 44(6): 95-103. doi: 10.13832/j.jnpe.2023.06.0095
Abstract(47) HTML (11) PDF(22)
Abstract:
For the reactor with closely arranged structure, it is particularly important to analyze the non-uniformity of circumferential temperature distribution of fuel rod cladding to prevent thermal stress failure of the cladding. In this paper, the theoretical analysis model is combined with the numerical fitting model, and the numerical fitting formula of the circumferential temperature distribution of the annular fuel rod in the lead-bismuth cooled tight cell is obtained under steady-state conditions, and compared with the numerical simulation data of computational fluid dynamics (CFD). The results show that the numerical fitting formula obtained in this paper has good accuracy, which can provide theoretical basis for the circumferential temperature analysis of fuel cladding of lead-bismuth reactor with annular fuel elements.
Experimental Investigation on Void Fraction Distribution of Boiling Flow in 2×2 Rod Bundle Channel
Liu Hao, Zhang Luteng, Zhou Wenxiong, Zhu Longxiang, Wan Lingfeng, Zhang Hong, Ma Zaiyong, Sun Wan, Pan Liangming, Deng Jiewen
2023, 44(6): 104-110. doi: 10.13832/j.jnpe.2023.06.0104
Abstract(43) HTML (8) PDF(23)
Abstract:
The conductivity probe method is an important methodology to obtain two-phase interfacial parameters. The radial distribution of the boiling flow void fraction in the 2 × 2 rod bundle channel was measured and analyzed by a one-sensor conductivity probe. The results show that under the current experimental conditions, from the fluctuation characteristics of the bubble signal, the fluctuation amplitude of the bubble signal main body at the center of the channel is small, and the fluctuation amplitude of the bubble signal main body at the rod gap is large. The radial distribution of void fraction shows the distribution characteristics of the core peak, and the core peak shows a concave phenomenon, which is due to the bubble at the center receives less heat than that at the wall.
Investigation on Characteristics of Resistance and Flow Distribution of 5×5 Annular Fuel Rod Bundle Channel under Pulsating Flow Condition
Li Jinyang, Ma Jun, Qiao Shouxu, Hao Sijia, Li Xupeng, Tan Sichao, Tian Ruifeng
2023, 44(6): 111-118. doi: 10.13832/j.jnpe.2023.06.0111
Abstract(53) HTML (13) PDF(17)
Abstract:
Compared to the traditional solid rod bundle channel, the annular fuel bundle channel has the advantages of enhancing cooling capacity and improving power density, and its special geometry makes the resistance characteristics closely related to the flow distribution characteristics. Under the pulsating flow condition, the flow rate in the inner and outer channels of annular fuel fluctuates periodically, which affects the heat exchange efficiency of coolant and threatens the safety of nuclear reactor. In this paper, a model of 5×5 annular fuel rod bundle channel is established based on computational fluid dynamics (CFD) method, and simulation calculation is carried out under steady-state and pulsating flow conditions. The simulation is benchmarked with the experimental velocity field measured with PIV and the friction factors predicted by the empirical correlation, and the results show good agreement. The variation characteristics of the friction factors of the inner and outer channels versus the Reynolds number are analyzed. Under the steady-state condition, the flow distribution ratio (inner channel flow versus outer channel flow) of the annular fuel is inversely proportional to the pressure drop ratio. Under the pulsating flow condition, the periodically averaged flow distribution ratio is inversely proportional to the pulsating frequency and proportional to the pulsating amplitude.
Distortion Evaluation of Natural Circulation Characteristic Curves Based on Scaled-down Methodology
Cheng Cheng, Lu Donghua, Su Qianhua, Zhang Ge
2023, 44(6): 119-126. doi: 10.13832/j.jnpe.2023.06.0119
Abstract(27) HTML (5) PDF(14)
Abstract:
Based on H2TS (Hierarchical, Two-Tiered Scaling) scaled-down method, three groups of experimental facilities, namely prototype, 1/2 model and 1/4 model, were designed and built. Natural circulation experiments on different heating power were carried out. The distortion of natural circulation characteristic curves in the drop-height scaled-down method was analyzed. The experimental results show that the parameters representing the main characteristics of natural circulation, such as natural circulation flow rate, Reynolds number, Grazoff number and modified Grazoff number, can keep the characteristic curves of natural circulation flow rate-heating power, Reynolds number-Grazoff number and Reynolds number-modified Grazoff number from being seriously distorted after being scaled down, and the fitting curves between the prototype and the model can form a one-to-one mapping. The characteristic curves of natural circulation can accurately reproduce the prototype law in the process of drop-height scaling. The reliability of drop-height scaling experiment simulating the natural circulation characteristic curves in the prototype is effectively verified.
Study on Transition Criteria of Slug Flow to Churn Flow in Vertical Rectangular Channel Based on Characteristics of Interfacial Force
He Qingche, Zhang Luteng, Pan Liangming, Xu Wangtao, Yan Meiyue, Sun Wan, Ma Zaiyong
2023, 44(6): 127-133. doi: 10.13832/j.jnpe.2023.06.0127
Abstract(30) HTML (9) PDF(12)
Abstract:
The transitional criteria between different flow regimes is of great significance to the safety analysis of nuclear reactor operation, because determining the flow regime of two-phase flow is a prerequisite for calculating the interfacial force and heat transfer of two-phase flow. In this study, four types of rectangular channels with cross-sectional size of 4×66 mm, 6×66 mm, 8×66 mm and 10×100 mm were employed to conduct the research on air-water upward flow in rectangular channels, and the void fraction and pressure drop were measured by impedance void meter and differential pressure gauge respectively. The results show that, at the constant superficial liquid velocity, the amplitude of frictional pressure drop reaches the maximum value at the transitional point of slug flow to churn flow, and when the void fraction is 0.6, the transition from slug flow to churn flow begins. The experimental results also show that the interfacial force gradient increases rapidly with the void fraction after the flow regime transition point, and the interfacial force gradient can be used as the criterion for the transition from slug flow to churn flow in terms of mechanical characteristics..
Experimental Research on Characteristics of Reactor Coolant Pump Torque under Two-Phase Conditions
Wang Kuo, Su Qianhua, Lu Donghua, Peng Fan, Xing Jun, Hong Rongkun
2023, 44(6): 134-139. doi: 10.13832/j.jnpe.2023.06.0134
Abstract(1148) HTML (4) PDF(16)
Abstract:
In order to grasp the torque characteristics of reactor coolant pump (RCP) under two-phase conditions, a special test platform was built in this study to carry out experimental research on a model pump based on the scale model of RCP. Firstly, the single-phase flow characteristics of the pump with multi-speed under normal condition, reverse pump condition, positive and negative energy consumption condition, obverse and reverse turbine condition, etc. were obtained. Then, the pump under single-phase air and two-phase 0.1-0.9 void fraction (the void fraction span is 0.1) was tested for the above conditions one by one. The proportionality law was used to sort out the experimental data at multiple speeds, and the variation rules of the torque under different void fractions, different flow rates and different conditions were obtained. Based on the data, homologous curve suitable for the primary circuit safety analysis software can be obtained simultaneously. The results show that with the increase of void fraction, the torque under the same condition decreases continuously, and the proportionality law curve drifts to zero gradually, while the homologous curve has the process of "degradation" first and then regression. When the void fraction reaches about 0.7, the degree of "degradation" reaches the maximum.
Nuclear Fuel and Reactor Structural Materials
Effect of Al Element on Thermal Aging Behavior of 20Cr25NiNb Heat-Resistant Steel
Shu Ming, Zhou Qin, Li Gang, Liu Xiao, Sun Yongduo, Zhao Ke, Xiao Jun
2023, 44(6): 140-147. doi: 10.13832/j.jnpe.2023.06.0140
Abstract(95) HTML (63) PDF(28)
Abstract:
In order to comprehensively investigate the high-temperature thermal aging behavior of supercritical gas-cooled reactor (SCGCR) cladding materials and the impact of Al element on the degradation of material mechanical properties, the thermal aging experiments at 750℃ were conducted on two types of 20Cr25NiNb austenitic heat-resistant steels: the alloy doped with Al and the Al-free one. Subsequently, corresponding microstructure analysis and mechanical property tests were carried out. The results revealed that the as-solutionized steels consisted of austenite along with a minor amount of micro-sized NbC carbides. After thermal aging, the matrix exhibited the precipitation of Laves and σ phases, while the alloy containing Al additionally showed the emergence of NiAl precipitates. The presence of Al element induced dual effects on the thermal aging behavior of 20Cr25NiNb. On one hand, Al element exhibited a solid solution strengthening effect and led to a reduction in size and an increase in number density of Laves particles after thermal aging, thereby enhancing high-temperature tensile strength. On the other hand, creep cracks predominantly initiated and propagated along grain boundaries. After thermal aging, the volume fraction of σ phase in Al steel was higher and the coarsening was more serious, consequently resulting in a notable reduction in creep fracture life. The fine Laves phase precipitated at grain boundaries in the Al-free alloy effectively suppressed the growth of σ phase, thus enhancing creep resistance. As a conclusion, this study offers robust support for the optimization of cladding material composition for SCGCR applications.
Study on the Thermal Aged Microstructure of Candidate Austenitic Heat-resistant Stainless Steel for Supercritical Water-cooled Reactor
Li Jun, Li Shaohong, Xiong Ru, Yang Hongmei, Li Mengnie
2023, 44(6): 148-154. doi: 10.13832/j.jnpe.2023.06.0148
Abstract(80) HTML (10) PDF(18)
Abstract:
In order to study the change of thermal aged microstructure and impact properties of alumina-forming austenitic stainless (AFA) steel, a candidate cladding material for Supercritical water-cooled reactor(SCWR), the AFA steel with 2.5% aluminum content was subjected to thermal aging treatment at 650℃ for 500~3000 h. The precipitated phases and the impact fracture were observed by field emission scanning electron microscopy. The types and crystal structures of the precipitated phases were studied by transmission electron microscopy. The results show that the impact toughness of the test steel decreases gradually with the extension of aging time, and the fracture of the test steel gradually transits from dimple fracture to mixed fracture mode of dimple fracture and cleavage fracture. The precipitation of Laves phase at grain boundaries and the precipitation and coarsening of γ'-Ni3Al phase during thermal aging are the main reasons for the decrease of impact toughness of AFA steel with the extension of aging time.
Research on Influence of Residual Pores on Thermal-Mechanical Performance of TRISO Particle in High Temperature Reactor
Zhao Yanli, Liu Shichao, Li Yuanming, Tang Changbing, Lu Huaiyu
2023, 44(6): 155-161. doi: 10.13832/j.jnpe.2023.06.0155
Abstract(28) HTML (7) PDF(9)
Abstract:
In order to investigate the influence of residual pores that may appear in SiC layer on the in-pile performance of TRistructural ISOtropic (TRISO) particle, and find the critical size of residual pores, in this paper, the in-pile performance of TRISO particle with residual pores was numerically simulated by using the multi-physical field coupling COMSOL software, and the effects of fission gas, CO release, internal pressure and residual pore size on the stress distribution of TRISO particle coating were analyzed. The results show that in the later stage of irradiation, the ratio of CO release is much higher than that of fission gas atoms, and the internal pressure of the particle can reach 49.5 MPa in the later stage. The existence of residual pores makes the stress of silicon carbide (SiC), inner dense pyrolytic carbon layer (IPyC) and outer dense pyrolytic carbon layer (OPyC) increase rapidly, especially in the SiC layer. When the size of residual pore reaches 9 μm, the maximum stress of SiC layer reaches 600 MPa, which is much higher than its intrinsic strength. When the residual pore size is 5 μm, the maximum stress of SiC layer is about 450 MPa, which is equivalent to its intrinsic strength. Therefore, in order to ensure the structural integrity of SiC layer in the preparation process, the residual pore size of SiC layer should be less than 5 μm.
Structural Mechanics and Safety Control
Performance Prediction and Structural Parameter Optimization of Control Rod Hydraulic Buffer Based on GA-BP Neural Network
Zhang Xiangwen, Fan Chenguang, He An, Wu Chuang, Yang Yujing
2023, 44(6): 162-169. doi: 10.13832/j.jnpe.2023.06.0162
Abstract(73) HTML (16) PDF(27)
Abstract:
In order to predict the buffer performance of hydraulic buffers of control rod assemblies by back-propagation (BP) neural network model improved by genetic algorithm (GA) and to realize the optimization of structural parameters. In this study, we simulated the falling rod in hydrostatic water for a specific control rod assembly hydrodynamic buffer. By changing the adjustable parameters of the test and setting up different test conditions, a large number of test data were obtained. The maximum impact force of control rod assembly in the process of rod falling was predicted by GA-BP neural network, and an optimized mathematical model was constructed. The nonlinear programming function (fmincon) is used to solve the problem, and a more optimal combination of structural parameters is obtained. The results show that the GA-BP neural network model has higher prediction accuracy compared with the BP neural network mdoel, and the fmincon function can realize fast solution of the optimal mathematical model of the maximum impact force of the control rod assembly. Therefore, the optimization method in this paper can provide some reference for the structural optimization design of hydraulic buffers.
Effect of Different Prestressed Conditions on Seismic Response of Nuclear Containment Structure under Ultimate Safety Ground Motion
Wang Xinxu, Li Xiaojun, Tang Hui
2023, 44(6): 170-178. doi: 10.13832/j.jnpe.2023.06.0170
Abstract(30) HTML (10) PDF(14)
Abstract:
In order to study the influence of prestressed conditions of concrete structure on the seismic response of the nuclear containment structure under ultimate safety ground motion, in this paper, only three prestressed conditions, which are instantaneous prestress loss, long-term prestress loss and uniform distribution of prestress after duct grouting, and the no prestressed condition are considered, and then the nonlinear time history analysis of the seismic response of the nuclear containment structure is conducted by establishing their refined finite element models. The simulation results show that for the no prestressed condition, the containment structure has a large number of cracks running through the thickness of the section in the weak seismic position under ultimate safety ground motion, and the structure basically enters the state of functional failure; while the whole containment structure is in an elastic state under the three prestressed conditions. However, cracks occur in the main opening areas of the containment, and a certain number of tension cracks are generated at the bottom of the containment body under the condition of uniform prestress distribution. The research shows that the prestressed concrete structure can significantly improve the seismic capacity of the nuclear containment structure, but it should be paid attention to the adverse impact of the uniform prestress distribution that may occur after the duct grouting on the seismic safety performance of the nuclear containment structure in design. The research results can provide guidance for the rational use of the seismic advantages of prestressed concrete in the design of nuclear containment structure.
Study on Seismic Decoupling Criteria of Primary and Secondary System of Nuclear Power Plant
Huang Yi, Luan Lin, Gao Fuhai, Qi Min, Ke Guotu, Li Xiaoxuan
2023, 44(6): 179-185. doi: 10.13832/j.jnpe.2023.06.0179
Abstract(42) HTML (16) PDF(13)
Abstract:
To expand the scope of seismic decoupling of primary and secondary systems in nuclear power plants, the requirements, basis, scope of application and limitations of seismic decoupling criteria in nuclear power plant codes were studied in this paper, and a decoupling model with modified stiffness was proposed. The coupling frequency of the primary and secondary system was derived by spring-mass model of two mass points, decoupling diagrams of each decoupling model were drawn based on the error analysis. Analyzed the applicability of each decoupling model, and compared seismic decoupling diagram criterion with seimic clause decoupling criterion. The results show that when frequency ratio of primary and secondary system λfi <1 and mass ratio λm <0.166, the decoupling model proposed in this paper can better simulate the secondary system frequency offset caused by coupling effect, with smaller frequency error; the seismic decoupling criteria can ensure that frequency errors of the primary and secondary system after decoupling are less than 10% when adopted decoupling model provided in codes. The proposed modified stiffness decoupling model expands the applicability range of the seismic decoupling diagram criterion. The research work provide references for the assessment of seismic decoupling and the selection of decoupling models for the primary and secondary systems of nuclear power plants.
Study on the Accident of SBO Superimposed by Steam-driven Auxiliary Feedwater Failure Based on RISMC Analysis Method
Wang Zhao, Li Qiongzhe, Guo Jianbing
2023, 44(6): 186-192. doi: 10.13832/j.jnpe.2023.06.0186
Abstract(39) HTML (6) PDF(12)
Abstract:
In order to study the safety performance of an in-service CPR1000 nuclear power unit under station blackout (SBO) accident superimposed by steam-driven auxiliary feed water failure, the risk-informed safety margin characterization (RISMC) analysis method is adopted; combining with accident scenario analysis, system reliability analysis, human factor engineering analysis and thermal hydraulic analysis, CARS software coupled with RELAP5 code is used to quantify the safety margin of the unit under accident. The safety performance and safety boundary of the unit under accident scenario is studied. The results show that the RISMC analysis method can effectively analyze the safety characteristics of units, and provide support for the operation and maintenance decision-making of nuclear power plants.
Research on Simulation Model of Small Integrated Lead-Bismuth Cooled Reactor
Sun Yuanli, Song Zhihao, Lyu Xiangbo
2023, 44(6): 193-198. doi: 10.13832/j.jnpe.2023.06.0193
Abstract(69) HTML (6) PDF(31)
Abstract:
Lead-bismuth cooled reactor has great potential in safety, design simplification, proliferation resistance and economic performance. In this paper, the small integrated lead-bismuth cooled reactor is taken as the research object. The helical-coil once-through steam generator (S/G) model based on four-equation drift-flux, the primary coolant system model, the constitutive model and the proportional-integral-differential (PID) control model are established, and the operation control characteristics of the lead-bismuth cooled reactor are studied. The results show that the steady-state calculation results are in good agreement with the design values, and the proposed model can accurately simulate the characteristics of the lead-bismuth cooled reactor. Under the condition of rapid load change, the system parameters overshoot is small, and the reactor power can follow the rapid change of steam flow. The blockage of heat transfer tubes has a great influence on the operation of the reactor, and the steam flow decreases by about 6.7% for each heat transfer tube blocked.
Analysis of Load-following Operation Capability of Large Sodium-cooled Fast Reactor
Dai Raoqi, Duan Tianying, Liu Yong, Zhang Houming, Zhang Weiying, Xu Weidong
2023, 44(6): 199-205. doi: 10.13832/j.jnpe.2023.06.0199
Abstract(46) HTML (15) PDF(23)
Abstract:
In order to study the load-following capability of a large sodium-cooled fast reactor in China, a simulation model of the whole system of this sodium-cooled fast reactor is established based on MATLAB/Simulink platform, and the load-following capability of the primary, second and third loops of the fast reactor system is tested in different levels under the extreme condition of load step and no control system intervention. The simulation results show that the primary and second loops of the fast reactor system can withstand ±10% of the load step, while the load-following capability of the third loop cannot meet this requirement. Under the current process design and key parameter limit conditions, this large sodium-cooled fast reactor has a poor load-following operation capability, and it cannot withstand a 10% load step. When operating at 90%Pn power level, it can withstand a 2.9% load step increase at most, and its once-through steam generator(OTSG) is the main limiting factor.
Analysis and Suppression of Load Oscillation in Underwater Fuel Assembly Transfer System
Yuan Zhanhang
2023, 44(6): 206-212. doi: 10.13832/j.jnpe.2023.06.0206
Abstract(21) HTML (5) PDF(6)
Abstract:
The stability of the underwater fuel assembly transfer system in nuclear power plant is essential to ensure the safety of the fuel assemblies loaded. Because of the flexibility of the steel wire rope, it is difficult to avoid the oscillation of the underwater running load trolley driven by the motor in the system. Considering the elasticity of the steel wire rope, a complete dynamic model of the underwater running load trolley was established. After decomposing the speed of the load trolley and simplifying it appropriately, the complete dynamic model was transformed into a simplified model, and the analysis and calibration of control system were carried out accordingly. A series notch filter was designed based on the simplified model to suppress the oscillation of the load trolley. The designed notch filter was applied to the complete dynamic model for simulation calculation. The simulation results showed that the oscillation of the load trolley was effectively suppressed, and the effectiveness of the simplified model was also verified. The analysis method in this paper can effectively suppress the oscillation of the load trolley during low-speed operation, which is simple, effective, and easy to implement in engineering.
Optimization Analysis of Physical Protection Design for Small Offshore Reactors
Liu Jian, Zhang Jiwei, Li Heng, Zhang Longqiang, Chen Huaping
2023, 44(6): 213-219. doi: 10.13832/j.jnpe.2023.06.0213
Abstract(30) HTML (8) PDF(19)
Abstract:
High economic efficiency is one of the important characteristics of small modular reactors (referred to as SMRs). Building a physical protection system for SMRs according to the current nuclear power plant physical protection system will have a negative impact on its economic efficiency. This article takes small offshore reactors as the analysis object, explores optimization schemes to reduce the cost of physical protection systems, and conducts two rounds of optimization and calculation demonstration for the schemes. Through analysis, although the physical barrier with high cost and on-site armed response forces have been eliminated for small offshore reactors, the effectiveness of the system can still be ensured by increasing the delay time of cabin passages. In the design of physical protection for SMRs, optimization should be carried out based on the characteristics of the reactor type, and a more cost-effective physical protection scheme should be selected while ensuring system effectiveness.
Circuit Equipment and Operation Maintenance
Research on Mobile Emergency Cooling System for Spent Fuel Pool in Nuclear Power Plant
Wang Zhixiao, Hu Jian, Peng Yue, Xie Enfei, Wan Qian
2023, 44(6): 220-225. doi: 10.13832/j.jnpe.2023.06.0220
Abstract(84) HTML (13) PDF(24)
Abstract:
In order to solve the issue of safe heat removal from the spent fuel pool after an accident similar to the Fukushima Daiichi Nuclear Power Plant Unit 4, a mobile emergency cooling system for the spent fuel pool in nuclear power plants has been developed by analyzing the characteristics of no power supply available on site, loss of cold chain, and emergency response time. The main equipment is centralized on the same mobile platform in the system, forming an integrated mobile emergency cooling device. The device comes with its own power source and cooling equipment that directly discharges heat into the atmosphere, with power self-sufficiency and a separate heat sink function. The device implements the modular design of cooling function, and each equipment in the device is fixedly connected. After an accident, it can be connected with the pre-laid water supply and return pipelines of the spent fuel pool through a hose. It can be quickly put into operation after an accident, and stably export the decay heat in the spent fuel pool. In addition, through the standardized design of the external dimensions of the mobile emergency cooling device, it can meet the relevant national standards for ordinary highway traffic inside and outside the plant. It can carry out mobile emergency response among different units in the plant, and can also carry out emergency rescue for units at other sites.
Research on Improvement of Pilot Solenoid Valve for Emergency Diesel Generator in Nuclear Power Plant
Fu Jiang, Zhang Lanqi, Tao Guoliang, Xiang Fangcheng, Liu Shengzhi, Chu Xiangnan, Wang Xiaojun, Yin Yulong
2023, 44(6): 226-230. doi: 10.13832/j.jnpe.2023.06.0226
Abstract(48) HTML (10) PDF(21)
Abstract:
The pilot solenoid valve used in the compressed air start-up system of emergency diesel generator in nuclear power plant fails frequently, and the failure modes are mainly manifested as inability to switch normally and jamming and air leakage. This failure will affect the reliability and availability of the compressed air start-up system of diesel generator and directly lead to the failure of diesel generator to start. In order to solve this problem, the defects in the structural design of the pilot solenoid valve were found through functional testing and disassembly, and it was clear that the root cause of the failure was the unreasonable reset structure of the pilot solenoid valve. The improvement measures mainly include the improvements of the main valve body structure, spool assembly interference structure, main spool guide structure, main spool return spring, diameter of the moving iron core crimping nozzle and the coil power of the pilot solenoid valve. The improved solenoid valve successfully completed all qualification tests. The test results show that the improved solenoid valve solves the failure of normal switching and jamming and air leakage, and improves the reliability of the air starting circuit of the emergency diesel generator.
A More Widely Applicable Calculation Method for Coastdown of Reactor Coolant Pump
Zhong Yun, Zhou Wenxia
2023, 44(6): 231-236. doi: 10.13832/j.jnpe.2023.06.0231
Abstract(34) HTML (9) PDF(9)
Abstract:
The applicable range of coastdown speed and flow analytical formulae based on similarity theory was investigated by using mature RELAP numerical calculation model, and then a more widely applicable calculation method for Reactor Coolant Pump (RCP) coastdown calculation was first proposed. The analytical formula of coastdown speed based on similarity theory is basically applicable to the conditions when the initial kinetic energy ratio (ε) is less than 3, while the analytical flow coastdown formula applicability is limited to conditions for which ε is lower than 0.3. When ε ≥ 0.3, the characteristics of flow, torque, head and ε deviate from similar conditions, which leads to invalid assumption of coastdown formulae, resulting in large error in calculation results of the formulae. Both the speed and flow coastdown formulae are applicable to coastdown calculation of PWR Reactor Coolant System and RCP test loops, because their ε is normally below 0.3; the error of coastdown formulae are within 10% normally, and the calculated speed error is less than flow error. A correction method for coastdown flow formula with ε ≥ 0.3 is proposed in this paper, which extends the applicable range of coastdown flow formula and is applicable for new reactor development.low formula with ε beyond 0.3 was developed, which expands the applicable range of coastdown formula and is applicable for new reactor development.
Research on Key Technology of Fuel Rod Oxide Film Thickness Measurement System
Xiao Xiang, Gao Sanjie, Gao Guangyong, Wen Junhao, Jiang Weiyu
2023, 44(6): 237-241. doi: 10.13832/j.jnpe.2023.06.0237
Abstract(649) HTML (14) PDF(30)
Abstract:
During the long-term service of nuclear reactor fuel assembly, zirconium alloy reacts with high-temperature water to form an oxide film. The measurement precision needs to reach ± 5μm due to the thin film. The high requirement of measurement precision brings difficulties to the development of the oxide film thickness measurement system. Therefore, based on the eddy current lift-off effect, a fuel rod oxide film thickness measurement system is developed in this paper. The key technology in the system is discussed, and the electromagnetic field of eddy current lift-off effect is simulated. The test results show that the measurement system has high precision, stability and reliability, and can measure the thickness of oxide film of in-service fuel rods.
Study on Off-design Operating Characteristics of Cold-end System for SCO2 Cycle Matching Fluoride-salt-cooled High-temperature Small Reactor
Zhao Quanbin, Zhao Kai, Chong Daotong, Liu Xiuting, Zhang Dalin, Zhuo Wenbin
2023, 44(6): 242-248. doi: 10.13832/j.jnpe.2023.06.0242
Abstract(624) HTML (5) PDF(11)
Abstract:
The small power generation system of supercritical carbon dioxide (SCO2) Brayton cycle coupled with fluoride-salt-cooled high-temperature small reactor is considered to have a good development prospect in the field of small nuclear power/nuclear power generation because of its high efficiency, compactness and high inherent safety. In this paper, the optimization design and off-design operation characteristics of cold-side for supercritical carbon dioxide Brayton cycle are studied. Based on the Fluoride-salt-cooled high-temperature small reactor characteristics and environment conditions, various SCO2 cycle configurations are compared, and it is found that the main compressor interstage cooling and reheat can improve thermal efficiency, but increase of limited, and extra heat exchangers are added. Then, the off-design operation characteristics of SCO2 cycle cold side system is investigated. The operation mode with constant pressure ratio not only has high cycle efficiency, but also has a wide adaptability range to the temperature change of the environment. Therefore, the constant pressure ratio operation mode is recommended for SCO2 cycle when the ambient temperature changes.
Analysis of Abnormal Rise of Fluorion in the Primary Circuit of PWR
Zhang Huazheng, Que Liangsheng, Zhang Jiakang
2023, 44(6): 249-253. doi: 10.13832/j.jnpe.2023.06.0249
Abstract(30) HTML (8) PDF(13)
Abstract:
In view of the abnormal increase of fluorion in the primary circuit of a PWR nuclear power plant after overhaul, in this paper, the interference factors of chemical analysis, the representative verification of sampling, reconnaissance of RCS makeup water quality, additives in the process system, impurities in fuel cladding, the use of chemical auxiliary materials, and water quality of related systems are investigated, the results show that the acid pickling passivation paste used for pickling passivation of stainless steel surface in the refueling pool and spent fuel pool during during construction and installation is not cleaned thoroughly and remains on the surface of stainless steel. The fluorion generated after irradiation decomposition pollutes the primary water, and causes the fluorion in the purification ion bed to be saturated, losing the purification ability of fluorion, and continuously releasing fluorion into the primary water during water exchange.
Study on Surface Passivation of 304L Stainless Steel under the Condition of Primary Water Chemistry in PWR Nuclear Power Plant
Cheng Wei, Jin Chengyi, Ji Dapeng, Wang Lei, Liu Hang, Wang Haochuan, Li Huaru
2023, 44(6): 254-259. doi: 10.13832/j.jnpe.2023.06.0254
Abstract(37) HTML (14) PDF(13)
Abstract:
The passivation reaction process of 304L stainless steel (304L SS) in the primary B+Li water chemistry environment of PWR nuclear power plant was simulated experimentally, and the effects of time and temperature changes on the passivation reaction process were studied. The results show that the oxide films grown formed by passivation on the surface of 304L SS has a double-layer structure, and the oxide films grown is in a state of continuous growth within 300 h, and the film thickness and corrosion resistance are positively correlated with passivation time. Increasing the passivation temperature can make the oxide films grown on the surface of 304L SS grow more evenly, thus improving the overall corrosion resistance of the oxide film on the surface. Based on the experimental results, this paper analyzes and puts forward a comprehensive growth model of oxide film, and discusses the engineering prospect of applying relevant conclusions to the first passivation test of hot functional testing.
Column of Science and Technology on Reactor System Design Technology Laboratory
Study on Loading Pattern for Long-cycle Lead-cooled Fast Reactor Core
Xia Bangyang, Xu Can, Qin Tianjiao, Li Qing
2023, 44(6): 260-265. doi: 10.13832/j.jnpe.2023.06.0260
Abstract(1748) HTML (28) PDF(70)
Abstract:
Lead-cooled fast reactor system is simple with natural safety, and it is one of the Generation Ⅳ nuclear power reactors with the most development potential and reality. Due to the high density, high operating temperature and opacity of lead coolant, the core refueling process is very difficult, complex and time-consuming, which affects the economy and safety of nuclear power plants. Therefor, increasing the core cycle length and reducing the refueling frequency have become an important aspect of the design and research of large-scale commercial lead-cooled fast reactor. Based on the engineering requirements, this paper analyzes the influence laws of the factors such as nuclear fuel type, fuel element and assembly form, reflector material, control rod layout, and the setting method of fissile nuclides blanket zone on the cycle length of lead-cooled fast reactor, and establishes the loading method of long-cycle core of lead-cooled fast reactor, which provides reference for the optimization of physical design of large commercial lead-cooled fast reactor.
Research on the Few Group Cross-section Production Method for Heat Pipe Micro Reactors Based on Monte Carlo Code
Xiao Peng, Luo Qi, Xia Bangyang, Yao Dong, Zhou Yajing, Fang Chao, Qin Tianjiao
2023, 44(6): 266-274. doi: 10.13832/j.jnpe.2023.06.0266
Abstract(58) HTML (12) PDF(19)
Abstract:
In order to improve the efficiency of the physical calculations of Heat Pipe Micro Reactors, based on the two-step method of cell homogenization-core transport calculation, this paper studies the few group cross-section production method for Heat Pipe Reactor core transport calculation by using the Monte Carlo (MC) code from the aspects of anisotropic scattering, fuel homogenization model, leakage correction and energy group structure. The numerical results show that using transport correction, leakage correction and other methods, and using a specialized "fuel-reflector" homogenization model for the peripheral fuel, the keff deviation obtained by the two-step core calculation is less than 100pcm (1pcm=10−5), the power distribution deviation is less than 3%, and the the control drum total worth deviation is less than 5%. Moreover, the core transport calculation cost is two orders of magnitude smaller than that of the Monte Carlo one-step full core calculation. Therefore, the two-step method of cell homogenization-core transport studied in this paper meets the accuracy requirements of engineering design and can greatly improve the efficiency of physical calculations of Heat Pipe Micro Reactors.