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2017 Vol. 38, No. 5

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Analysis for Startup Physics Test of First 18 Months Refueling for Unit 1 of Hongyanhe Nuclear Power Plant
Zhang Haizhou, Cao Yunlong
2017, 38(5): 1-3. doi: 10.13832/j.jnpe.2017.05.0001
Abstract(13) PDF(0)
Abstract:
The start-up physical test of 18-months refueling for unit 1 of Hongyanhe Nuclear Power Plant has completed. The results show that the predicted value of 18-months refueling theory is in good agreement with the measured results. The results verified the accuracy of refueling design. In this paper, the results from the startup physical tests after 18 months refueling and the 12 months refueling were compared, and the changes of core characteristics for 18-months refueling core are pointed out and analyzed.
PIV Experiment Research on Flow Pattern in Secondary Side of Steam Generator Based on a Visible Experimental Facility
Wang Cong, Lu Daogang, Yao Zhipeng, Cao Qiong, Awais Ahmad, Zhang Shuming
2017, 38(5): 4-9. doi: 10.13832/j.jnpe.2017.05.0004
Abstract:
The thermal-hydraulic characteristics of secondary side has been researched based on this facility. And a two-dimensional particle image velocimetry(2D-PIV) experiment was performed to obtain the flow pattern in the secondary side. We found that the transverse flow phenomenon occurred in the vertical tube region. The experiment also got the gradient of velocity field under different power and varying ratio of feedwater flow rate on hot side and cold side. Less intense transverse flow was observed in the case of non-uniform ratio of feedwater at both legs as compared to uniform flow.
Experimental Research of Natural Circulation Characteristics in Secondary Side Passive Residual Heat Removal System
Xi Zhao, Sun Doucheng, Zhu Yuan, XiE Feng, Li Yong, Zan Yuanfeng, Zhuo Wenbin
2017, 38(5): 10-13. doi: 10.13832/j.jnpe.2017.05.0010
Abstract(15) PDF(0)
Abstract:
The natural circulation characteristics in the secondary side passive residual heat removal system is experimentally studied on ESPRIT(Emergency Secondary Passive Residual heat removal system Integral Test facility). The experiment is carried out under prototypical, elevated power and elevated resistance conditions, and the response characteristics and operation capability were studied under station blackout scenario. The test data reveal that the design volume of the cooling water tank satisfies the heat removal demand in 72 hours. The heat sink possesses sufficient cooling capability after the power is elevated by 6%. The system pressure in elevated resistance condition is always higher than that of the prototypical condition. The natural circulation always remains and the system temperature and pressure keep decreasing in elevated resistance condition.
Experimental Study of the Effect of Original IRWST Temperature on PRHRS Operation Characteristics
Huang Zhigang, Zhang Yan, Peng Chuanxin, BAi Xuesong, Zhuo Wenbin, Yan Xiao
2017, 38(5): 14-17. doi: 10.13832/j.jnpe.2017.05.0014
Abstract(10) PDF(0)
Abstract:
The PRHRS operation characteristics have been experimentally studied under different initial IRWST temperature conditions. The experimental results are compared in this paper, such as core inlet/outlet temperature, system pressure, nature circulation flow rate of PRHRS, heat exchange power. The comparison of the experimental results showed that the lower initial IRWST temperature caused the core inlet/outlet temperature and system pressure drop faster, the PRHR HX outlet temperature is lower but the heat exchange power is higher, the trends of nature circulation flow rate of PRHRS is similar.
Development and Verification of Transport Code for Complicated Geometry Based on CSG Formula and OpenMP
Zheng Yong, Peng Minjun
2017, 38(5): 18-23. doi: 10.13832/j.jnpe.2017.05.0018
Abstract(19) PDF(0)
Abstract:
The small research reactor which exhibits the features of high heterogeneity, strong absorption regions and complicated geometry challenges the numerical computation capabilities of existing neutronics codes. To overcome the restriction of geometry, the transport code MOCAGE, based on the constructive solid geometry formulation(CSG) and matrix method of characteristics, has been developed in the present study, and the track tracing has been parallelized applying the OpenMP programming model. The irregular geometry problem and HTTR benchmark problem with three configurations of control rods has been used to verify the capability of characteristic line tracing and to evaluate the code’s numerical accuracy for complicated geometry. The system effective multiplication factor, the normalized fission densities of fuel pins, and the normalized neutron absorption densities of burnable poison rods and control rods have been calculated by the developed code. Meanwhile the comparison with MCNP5 reference solutions has been performed, and the results demonstrated that the developed code can model the complicated geometry and trace the characteristic lines correctly, and excellent agreements were achieved with respect to the aforementioned parameters. The time consumption of tracing procedure has been reduced significantly using the OpenMP.
Development of A Nodal Model for Space Reactor System Dynamic Characteristics Analysis
Li Huaqi, Hu Pan, Yang Ning, Zhu Lei, Tian Xiaoyan, Chen Lixin, Jiang Xinbiao
2017, 38(5): 24-27. doi: 10.13832/j.jnpe.2017.05.0024
Abstract(13) PDF(0)
Abstract:
A lumped parameter method nodal model is developed for alkali metal cooled space reactor system dynamic characteristics analysis in this paper. The space reactor system dynamic response code was built by using Simulink, and the code was approved with steady state designed parameters. The transient response of space reactor was studied on the step responses of the increase in the drum angle and external load. The results show that the power increased rapidly, and then the power increased slowly to a final new steady state because of the core reactivity temperature feedback. When increased in external load, the TE electric power output increased rapidly and a great negative reactivity was brought to the core, thus the core underwent a rapid power decrease and core fuel temperature decreased. The TE electric power output responds to the change in external load in a much faster spend than the change in control angle.
Investigation of RELAP5 Modeling Method of Single Phase Reversed Flow in Inverted U-Tube of Steam Generator
SHen Mengsi, Yu Lei, Hao Jianli, Hu Gaojie
2017, 38(5): 28-33. doi: 10.13832/j.jnpe.2017.05.0028
Abstract:
Using the code RALAP5/MOD3.2 to model and calculate the experiment of single phase reversed flow of steam generator, the results show that it is inaccurate to calculate the reversed mass flow rate using the model, which models the inverted U-tube by its length, and the new model based on it can more accurately calculate the reversed mass flow rate. Lastly, a method applied to investigate the single phase reversed flow in inverted U-tube of steam generator of nuclear plant is put forward.
Development of Code for Steady-State Thermal-Hydraulic Analysis in Bimodal Space Nuclear Reactor with Heat Pipe
Tian Xiaoyan, Jiang Xinbiao, Chen Lixin, Li Huaqi, Yang Ning, Zhu Lei, MA Tengyue
2017, 38(5): 34-39. doi: 10.13832/j.jnpe.2017.05.0034
Abstract(15) PDF(0)
Abstract:
In order to study the core steady-state safety characteristics of Bimodal Space Nuclear Reactor with Heat Pipe( HP-BSNR), the thermal-hydraulic models of modified core are established and a steady-state thermal-hydraulic analysis code for the HP-BSNR named STHAHPBSNR is developed based on the preliminary conceptual design of HP-BSNR. The hydrogen property and steady-state thermal-hydraulic parameters calculated by STHAHPBSNR are compared with that calculated by program ELM as well as the experimental data in the published literatures, which turn out to agree well. Besides, the effect of different heat transfer and friction resistance correlations on the channel wall temperature is studied. It demonstrates that the STHAHPBSNR can provide the initial steady-state parameters for the transient-state safety analysis of HP-BSNR.
Simulation Research on PZR Water Seal of Nuclear Power Plants
Fu Guanhua, Li Quanbing, Ren Hongbing, Zhou Peng, Duan Yuangang
2017, 38(5): 40-44. doi: 10.13832/j.jnpe.2017.05.0040
Abstract(14) PDF(0)
Abstract:
A computational fluid dynamics(CFD) model of vapor-liquid with non-condensing gas is established, and Fluent is used to simulate the flow field in PZR water seal of a nuclear power plant. The thermal characteristics of water seal formation are visualized. The effects of PZR pressure and non-condensing gas on water seal formation are researched. The results show that water seal formation time is shortened with the increasing of pressure, and is prolonged with the increasing of non-condensing gas content.
Study on Structural Integrity of Main Pipeline of Advanced PWR Nuclear Power Plant
Chu Qibao, Fang Yonggang, Wang Qing, Nan Xiangchen
2017, 38(5): 45-48. doi: 10.13832/j.jnpe.2017.05.0045
Abstract(15) PDF(0)
Abstract:
Taking a nuclear power plant as an example, the structural integrity of the nuclear safety class 1 pipe is analyzed and evaluated in this paper, and the design margin of the pipeline is analyzed according to the standard. The structural strength of pipeline is evaluated according to the standard, thermal ratchet is evaluated by analyzing the pipeline temperature theoretically, and the fatigue life of the pipeline is evaluated by using the simplified rain flow method. The results show that the minimum wall thickness of the main coolant pipe meets the requirements of the standards when reduced to 55 mm, though the safety margin is small. The safety margin of fatigue and thermal ratcheting evaluation of branch is minimum.
Comparison of Leak-Before-Break Assessment of Main Loop Piping Lines Fabricated of Different Materials
MA Linwei, HE Jiasheng, Shu Anqing, ZhenG Xiaotao, Xu Jianmin, Yu Jiuyang
2017, 38(5): 49-53. doi: 10.13832/j.jnpe.2017.05.0049
Abstract(11) PDF(0)
Abstract:
As to the LBB evaluation performed in China, the materials are cast austenitic stainless steel(CASS) or wrought stainless steel(WSS). In this paper, LBB assessment in the guidance of SRP 3.6.3 was performed to evaluate the main loop piping lines of CASS and WSS. The adjustment due to thermal aging is performed to achieve reasonable material properties. J integral/tearing modulus approach is used to determine the critical crack size of CASS pipe and limit load approach is used to determine the critical crack size of WSS pipe. Leakage flaw size is determined based on Henry’s homogeneous non-equilibrium critical flow model. In order to demonstrate that fatigue crack growth is not a potential source of pipe rupture for the evaluated piping lines, the fatigue crack growth of a postulated circumferential part-through-wall crack and the fatigue crack growth of a circumferential through-wall crack are analyzed. The results show that WSS material is better than CASS material in LBB capability.
Method for Calculation of Fretting Wear of PWR Fuel Rod Cladding
Qi Huanhuan, Feng Zhipeng, Wu Wanjun, Jiang Naibin, Huang Xuan
2017, 38(5): 54-57. doi: 10.13832/j.jnpe.2017.05.0054
Abstract(14) PDF(0)
Abstract:
Archard wear formula was used as the theoretical model for fretting wear of PWR fuel rod cladding, and the fretting wear volume between fuel rod cladding and grid would be predicted through this formula, where the significant physical quantities were wear coefficient, contact force between fuel rod and grid, and sliding distance. Wear coefficient was determined by experiment. Contact force between fuel rod and grid was a function varied with assembly burnup, which was determined by experiment or engineering empirical formula. The maximum vibration displacement for all modes would produce relative sliding if the displacement exceeded the threshold which was defined by grid dimple stiffness, contact force between fuel rod and grid, and friction between them, and the sliding distance could be estimated in an infinitesimal time increment. After the three physical quantities were determined, wear formula integral was implemented to obtain the fretting wear volume. According to the wear geometry of cylinder-plane contact, the relationship between wear volume and depth was derived theoretically, then the wear depth could be obtained from the wear volume. Finally comparing the wear depth to the criterion, it was validated that whether the fuel rod could satisfy the requirement on mechanism integrality.
Analysis and Discussion for Fast Fracture of Reactor Pressure Vessel (RPV) Based on RCC-M Code
Wang Dasheng, Liu Pan, JiN Ting, Lu Wenjie
2017, 38(5): 58-61. doi: 10.13832/j.jnpe.2017.05.0058
Abstract(14) PDF(0)
Abstract:
The fast fracture analysis and evaluation methods were compared and discussed in appendix ZG of RCC-M code Ver.2000 and Ver.2007. Comparison shows that the appendix ZG 2007 edition has a wider scope of application, the process of analysis in Ver.2007 is relatively simplified and the consideration in Ver.2007 are more comprehensive for the material aging. Taking the analysis of the fast fracture of the reactor core as an example, the sensitivity of the orientation parameters of the reference defects is analyzed, and the evaluation results of the two versions were compared. The results show that the margin of the evaluation results of Ver.2007 appendix ZG is much larger and the limitation of the P-T curve of the reactor pressure vessel is enlarged, and the operation space of the nuclear power plant is expanded
Buffer Test and Analysis of CENTER HFETR Control Rod Driver Mechanism
Wu Xiaofei, Li Shuo, Nie Changhua, Yang Zumao, Yan Xiao, Wang Xiaoheng, Xing Limiao
2017, 38(5): 62-66. doi: 10.13832/j.jnpe.2017.05.0062
Abstract(10) PDF(0)
Abstract:
The research reactor CENTER HFETR is with frequent start-up, shut-down and power follow-up, and the designing of the CRDM should focus on the buffering effect to ensure its reliability during its life time. Buffering tests are conducted on six models, and the buffer structure is selected based on the comparison of the control rod dropping impact and dropping time in the condition of static and dynamic water. The selected structure is with proper dropping time and small impact, which is applicable for this research reactor.
Critical Buckling Load Calculation Method of Nuclear Grade Dynamic Tension Bars
He Mengfu, Liu Karen, Han Lang, Cao Leisheng
2017, 38(5): 67-71. doi: 10.13832/j.jnpe.2017.05.0067
Abstract(14) PDF(0)
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In this paper, the critical buckling loads of these non-uniform section bars are checked by using a quick and effective theoretical calculation method to meet the engineering application. Transfer matrix method used in this method is checked by using the finite element and KTA3205 test method. The differences of these results are analyzed to verify that the theoretical calculation is reasonable and usable by comparing with the manufacturing process and the simplified calculation method.
Comparative Analysis of Timeliness and Reliability of OBE Alarm Strategies in Nuclear Power Plants
Chen Zhigao, Fan Tao, Lu Jianqi, Li Shanyou
2017, 38(5): 72-76. doi: 10.13832/j.jnpe.2017.05.0072
Abstract(12) PDF(0)
Abstract:
Based on 1642 records of 31 earthquakes in China from 2003 to 2015, the timeliness and reliability of OBE alarm strategies used in 2nd and 3rd GNP were compared and analyzed. The results showed that, for the same ground motion, when OBE was beyond limit for both 2nd and 3rdGNP, the OBE alarm of 3rd GNP were later than that of 2nd GNP, with an average delay of 20 seconds. Peak ground acceleration could be as high as 0.7 g before 3rd GNP alarmed. OBE over-limit determination performed using data of 108 stations from the region of Ⅵ intensity of 4 destructive earthquakes showed that the proportion of stations exceeded limit among 3rd GNP, 2ndGNP(CGN), and 2nd GNP(CNNC) were 41%, 63% and 76%. It indicated that due to the aim of lowering the rate of unnecessary shutdown caused by near-field earthquakes with small magnitude and large amplitude pulses and earthquakes with small amplitude and long duration ground motions, two-parameter decision method was applied in threshold selection, thus the timeliness and reliability of 3rd GNP was reduced, which is insecure for important nuclear facilities.
Design of Safety Logic Channel Periodic Test Device for HFETR
Wu Wenchao, Li Linhong, Li Pu, Chen Qibing, Li Ziyan
2017, 38(5): 77-80. doi: 10.13832/j.jnpe.2017.05.0077
Abstract(13) PDF(0)
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The safety logic channel periodic test device for high flux engineering test reactor(HFETR) was designed to detect the availability of the safety logic channel, and to find the reject-action fault of the safety logic channel. The operation principle and equipment composition of the device were introduced. The device was tested with manual and automatic modes. The effects of the test device on the logic channel were analyzed. Long-term operation practice shows that the safety logic channel periodic test device meets the test demand.
Improvement of Generating Branch for DDET——Accurate Probabilistic Threshold Value Method
Guo Haikuan, Zhao Xinwen, CAi Qi, Zhang Yongfa, Huang Liqin
2017, 38(5): 81-85. doi: 10.13832/j.jnpe.2017.05.0081
Abstract(14) PDF(0)
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Probabilistic threshold value method is often used to generate the branch for dynamic discrete event tree, which is with two deficiencies: inaccurate time of generating branch and time uncertainty for establishing of dynamic discrete event tree. This paper presents an accurate probabilistic threshold value method to overcome the deficiencies. The analysis of the dynamic cistern finds that, when the probabilistic threshold value is 0.99, with the accurate probabilistic threshold value method, the total nodes are 25 % less, the accident node are 2.7% less, and the counting time is 37.7% less that obtained by the probabilistic threshold value method. It shows that the accurate probabilistic threshold value method is with higher analytic efficiency of accident node than the probabilistic threshold value method, and the accident probability obtained by the accurate probabilistic threshold value method is 33.6% less than that obtained by the probabilistic threshold value method. In a dry out accident, it gives the operation staff longer time to deal with the accident; and the dynamic discrete event tree established by the accurate probabilistic threshold value method shows more evolution paths and accidental information, which is helpful for the operation staff to fully understand the evolution of dynamic cistern accident and suitable for the study of the system operational dynamic characteristics.
Design of a Removable Emergency Power Supply System for HFETR
Tan Fujun, Li Changshun, Jin Yang, Xu Chuan, Qin Yongcheng
2017, 38(5): 86-90. doi: 10.13832/j.jnpe.2017.05.0086
Abstract(10) PDF(0)
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When a large earthquake occurs, it is highly possible that all of the power supply systems of the nuclear reactor fail, and it is a huge menace for the security of nuclear reactor. In order to enhance the defense in depth of High-Flux Engineering Test Reactor(HFETR), a removable emergency power supply system(REPSS) is designed as the emergency rescue facilities of HFETR, which is independent of the main power supply system, removable, fast to connect or excide. The paper calculates and analyzes the design capacity of REPSS, and briefly narrates the technical requirements, structure plan and qualification method of anti-seismic diesel generator set of REPSS.
Reliability and Parameter Sensitivity Analysis of Passive System in Nuclear Power Plants
Jiang Lizhi, Cai Qi, Zhang Yongfa
2017, 38(5): 91-95. doi: 10.13832/j.jnpe.2017.05.0091
Abstract(12) PDF(0)
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Under the framework of RMPS(Reliability Methods for Passive Safety Functions), thermal-hydraulic reliability(TH-R) of an experimental facility is evaluated, and sensitivity analysis is applied for uncertainty parameters which has effects on the TH-R of the passive systems. Related conclusions contribute to understand the uncertainty of the thermal hydraulic processes of the passive residual system from the view of reliability. Results of reliability and sensitivity analysis can be used to guide the system optimal design and operation management.
Development of a Coordinated Control Strategy for Space Reactor Based on PI Controller
Li Huaqi, Hu Pan, Zhu Lei, Yang Ning, Tian Xiaoyan, Ma Tengyue, Chen Sen, Jiang Xinbiao, Chen Lixin
2017, 38(5): 96-100. doi: 10.13832/j.jnpe.2017.05.0096
Abstract(11) PDF(0)
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A proportional-integral(PI) control of space reactor electric power method is developed. The control scheme of electric power by manipulating the reactivity only or external load only was analyzed. The result shows that the electric power responds at a slower speed by manipulating the reactivity only. When manipulating the external load only, the electric power will be out of the control range and the system becomes unstable. Thus, an optimal PI controller of the space reactor electric power that based on coordinated control strategy on the external load and the reactivity was developed. The result shows that the coordinated control strategy can avoid the limitations of the control of electric power by manipulating the reactivity only or external load only. The coordinated control strategy can meet the basic control requirements of the space reactor electric power output.
Study on Maximum Speed Setting Standard during Speed up Process of Reactor Coolant Pump
Su Songzhou, Wang Pengfei, Xu Zhongbin, Ruan Xiaodong, Kong Weijie
2017, 38(5): 101-105. doi: 10.13832/j.jnpe.2017.05.0101
Abstract(14) PDF(0)
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In order to get the maximum speed setting standard during the temperature-rise period of CAP1400 reactor coolant pump which rated speed was 1500 rpm, the third similarity law and dimensional analysis were used to derive the setting formula of rotation speed which guaranteed that the axial force and power did not exceed the value of normal operating point. The maximum speed of AP1000 and CAP1400 were obtained according the parameters of the normal operating point, and the calculated value of AP1000 was 87.5% rated speed, which was corresponded to the current maximum speed setting of AP1000. The validated numerical computation of the self-designed RCP hydraulic model was carried out to verify the axial and radial force in different settings. The results show that the relation of axial force conforms to the theory when the maximum speed was set to ensure the safety of the axial force. The setting formula of the rotation speed which guaranteed both axial force and power can be used to set the maximum speed during startup process of CAP1400 reactor coolant pump.
Investigation and Application of Reactor Fuel Assembly Dismantling Technology
Liu Xiaosong, Li Wenyu
2017, 38(5): 106-109. doi: 10.13832/j.jnpe.2017.05.0106
Abstract(10) PDF(0)
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By analyzing the reactor fuel assembly structure, a fuel assembly dismantling process is obtained, which includes overturning, removing the tube socket and finally extracting the fuel rods. This process can be adapted for narrow space and is simple for operation and safe. To accomplish those operation, an appropriate tool with the projected area less than 1.2 m~2 had been developed by an integrated design approach. It meets the requirement of on-site condition and is good for preserving these existing devices. Through this study and application, the reactor fuel assembly has been disintegrated based on this technique under water in the depth of 6 m.
Air Lift System Based on PROFIBUS DP Mass Flow Controller for Nuclear Application
Zhang Bo, Wu Ke, Chen Chaodong, Li Xiaowei
2017, 38(5): 110-114. doi: 10.13832/j.jnpe.2017.05.0110
Abstract(11) PDF(0)
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The control method of mass flow controller based on PROFIBUS DP is presented for the air lift system for nuclear fuel reprocessing. Based on the introduction of the composition of air lift system, this paper presents the structure of the control system, and studies the selection of the PROFIBUS DP fieldbus mass flow controller, the electric connection and the software realization. Finally, the relationship between the compressed air flow and the liquid flow is established by the polynomial regression method.
Preliminary Study on Evaluation Criterion of CPR1000 Residual Heat Removal Pump Operability upon Seismic Condition
Chen Xingjiang, SHao Chunbing, Yang Jinchun, Huang Qiongyu, Cong Guohui
2017, 38(5): 115-118. doi: 10.13832/j.jnpe.2017.05.0115
Abstract:
The operability of nuclear safety class pump upon seismic condition influences the safety of nuclear power plants directly. The function of pump upon seismic condition should be qualified. The CPR1000 motor-driven residual heat removal pump is used to act as subject investigated in this paper. According to the result of the failure mode and effect analysis(FMEA), the weakness of pump upon seismic condition which affects the operability is selected. Based on the function and operating requirement of the assembly, the evaluation criterion of pump operability upon seismic condition is provided in the paper.
Analysis of Physical and Thermal Characteristics of Damaged Fuel Assemblies after Repairment
Chen Qiuyang, XuE Feng, Gao Yongjun
2017, 38(5): 119-122. doi: 10.13832/j.jnpe.2017.05.0119
Abstract(14) PDF(0)
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This report takes CPR1000 unit with AFA3 G fuel assembly as the research object, and analyzes the physical characteristics and power distribution of normal fuel assembly and damaged fuel assembly after repairment. It can be found that the effect on the fuel assembly reactivity is about-0.03% for the fuel assembly replaced with a stainless steel rod, and the effect can be ignored. The relative power of fuel rods near the stainless steel rod increases about 5.6%. The stainless steel rod replaced in the fuel assembly has effect on the relative power of fuel assembly about 0.1372% to 0.2698%, on burnup about 0.11%, core moderator temperature coefficient about 2%, and these effects can be ignored. The stainless steel rod replaced in the fuel assembly has no effect on the reactor core power peak factor, core critical boron concentration, core shutdown margin, fuel assembly outlet moderator temperature and core outlet coolant temperature.
Study on Corrosion Resistance of Zr-0.8Sn-1Nb-0.3Fe Alloy after Kr+ Ion Irradiation
Yang Zhongbo, Cheng Zhuqing, Qiu Jun, Wu Zongpei, Zhang Hai, Ran Guang
2017, 38(5): 123-128. doi: 10.13832/j.jnpe.2017.05.0123
Abstract(16) PDF(0)
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The corrosion resistance of Zr-0.8 Sn-1 Nb-0.3 Fe alloys prepared by two different processes was investigated in 400℃/18.6 MPa superheated steam by static autoclave after irradiated by 360℃ with Kr+-irradiation of 5~25 dpa. The microstructures of oxidation film after corrosion were analyzed by TEM, SEM, and XRD. The results showed that the corrosion weight-gain increased with the irradiation dose, while the weight-gain of 1# alloy with smaller and more dispersive SPPs than 2# alloy was lower under the same irradiation dose. Before corrosion turning, the oxygen content in the oxidation film decreased from the steam-side to the zirconium matrix. The oxidation film beside the steam-side was mainly composed by equiaxied monoclinic ZrO2 crystal, while near the film/matrix interface by columnar quartet ZrO2 crystal and hexagonal Zr3O crystal. After transition of corrosion weight, the film near the interface grew like cauliflowers, and the size of cauliflowers were corresponded to the growth rate and uneven growth trend of oxidation film.
Study on High-Cycle Fatigue Behavior for TA16 and 690 Tube
Zhao Yuxiang, Liu Ranchao, He Kun, Xiong Ru, Wang Li
2017, 38(5): 129-131. doi: 10.13832/j.jnpe.2017.05.0129
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The high-cycle fatigue experiments for steam generator tube including TA16 and 690 were conducted under axial loadings. The environments were at room temperature in air. The S-N curves were processed according to the experimental data, and the fatigue limit of the cycle is obtained by fitting formula. The fracture morphology was observed by SEM, and the fracture morphology presents areas of crack initiation, crack growth and fracture.
Study on Corrosion Resistance of Zr-Sn-Nb-Fe Zirconium Alloys
Cheng Zhuqing, Yang Zhongbo, Qiu Jun, Liu Hong, Yuan Gaihuan, Gao Bo
2017, 38(5): 132-137. doi: 10.13832/j.jnpe.2017.05.0132
Abstract(16) PDF(0)
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The corrosion resistance of stress relieved and recrystallized SZA-4(Zr-0.8 Sn-0.25 Nb-0.35 Fe-0.1 Cr-0.05 Ge) and SZA-6(Zr-0.5 Sn-0.5 Nb-0.3 Fe-0.015 Si) alloys, and stress relieved reference alloy A(Zr-1 Sn-1 Nb-0.1 Fe) are studied by static autoclaves in three different water conditions: 360℃/18.6 MPa deionized water, 360℃/18.6 MPa/0.01 mol·L-1 lithiated water and 400 ℃/10.3 MPa superheated steam. SEM and TEM are used to analyze the microstructures. The results showed that the corrosion resistance of SZA-4 and SZA-6 are both better than that of alloy A in all the water conditions. The corrosion resistance of recrystallized SZA-4 is better than that of stress relieved SZA-4, while SZA-6 shows the reverse phenomenon. There are two kinds of SPPs in SZA-4, one is Zr(NbFeCr)2 with a smaller size and the other is Zr(NbFeCrGe)2 with a larger size, and both have a HCP structure. In SZA-6, there exists(ZrNb)2 Fe SPPs with a FCC structure and Zr(NbFe)2 SPPs with a HCP structure. It is concluded that the chemical compositions and the characteristics of SPPs are responsible for the difference of the corrosion resistance, and the former plays a dominant role.
Evaluation on I-SCC Properties of Zirconium Cladding
Yan Meng, Wang Pengfei, Hong Xiaofeng, Liang Bo, DAi Xun
2017, 38(5): 138-140. doi: 10.13832/j.jnpe.2017.05.0138
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Hoop tensile test were performed on zirconium(such as N36, Zr-4 and X) ring specimens at 350℃ and 400℃ to study the iodine induced stress corrosion cracking(I-SCC) behavior of zirconium cladding in 10~2 Pa, 10~3 Pa and 10~4 Pa iodine partial pressure environment. Different levels of I-SCC happened with N36, Zr-4 and X samples at special iodine partial pressure with the maximum load, and the fracture energy dropped down quickly as result, and N36 maximum load and fracture energy dropped slower at same test situation.
Research on Roller Status Diagnosis of CRDM Based on Simulation Method
Yang Xiaochen, Li Wei, Zhang Liming, Yang Fangliang, Zhang Zhifeng
2017, 38(5): 141-144. doi: 10.13832/j.jnpe.2017.05.0141
Abstract(13) PDF(0)
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Abrasion is the primary cause for the failure of the roller screw drive type CRDM. This paper describes a method of CRDM roller state diagnosis based on computer simulation. Through a common simulation software ADAMS, the vibration characteristics of the roller under normal conditions, pit defects and excessive abrasion have been obtained. Research indicates that the roller will have a significant impact signal while working. With the deterioration of conditions, the impact signal will be enhanced.
Coupled 3-D Neutronics/Thermal-Hydraulics Analysis for SCWR Core Typical Transients
Wang Lianjie, Zhao Wenbo, Chen Bingde, Yao Dong, Lu Di
2017, 38(5): 145-150. doi: 10.13832/j.jnpe.2017.05.0145
Abstract(11) PDF(0)
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Transient performance of CSR1000 core during some typical transients, such as CR ejection and uncontrolled CR withdrawal is analyzed and evaluated with the coupled three dimensional neutronics/thermal-hydraulics SCWR transient analysis code. The 3-D transient analysis shows that the maximum cladding surface temperature retains lower than safety criteria 1260℃ during the process of CR ejection accident, and the maximum cladding surface temperature retains lower than safety criteria 850℃ during the process of uncontrolled CR withdrawal transient. The safety of CSR1000 core can be ensured during the typical transients under the salient fuel temperature and water density reactivity feedback and the essential reactor protection system.
Experimental Investigation on Flooding of AP1000 Pressurizer Surge Line
Tian Wenxi, Yu Jiangtao, Wang Zhiwei, Su Guanghui, Qiu Suizheng
2017, 38(5): 151-155. doi: 10.13832/j.jnpe.2017.05.0151
Abstract(10) PDF(0)
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With air-water as working fluid, the tests of flooding in AP1000 surge line are conducted at atmospheric pressure and room temperature with different dimensional water levels in the pressurizer simulator. Mechanism analysis is performed on the flooding phenomena and data, and their dynamic nature has been captured. The research result shows that, flooding is most likely to transpire on the vertical part of the surge line, flooding characteristic coincides with the Kutateladze relation, the water flow rate decreases as the pressurizer simulator water level increasing in small gas flow conditions, the water flow rate increases as the pressurizer simulator water level increasing in large gas flow conditions, and zero water penetration point is almost independent of the pressurizer simulator water level.
Development of a Model for Pure Steam Condensation in a Vertical Tube
Gou Junli, Wang Baojing, Ding Wenjie, Dan Jianqiang
2017, 38(5): 156-159. doi: 10.13832/j.jnpe.2017.05.0156
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In order to improve the capability of thermal-hydraulic codes to predict the heat transfer coefficient(HTC), an analytical model is developed based on the conservation equations. The calculated values of this model and of Relap5 were compared with the Kuhn’s experimental data. It is found that the model gives better prediction of HTC than Relap5, and it agrees well with the experimental data.
Heat Transfer Simulation of Fuel Transport Cask for High Temperature Gas Cooled Reactor
Liu Yang, Wang Jun
2017, 38(5): 160-163. doi: 10.13832/j.jnpe.2017.05.0160
Abstract(14) PDF(0)
Abstract:
The heat transfer process(conduction, convection, and radiation) of the transport cask for the high temperature gas cooled reactor is simulated by CFX code. The calculation shows that every component of the cask is below its limit temperature, so the thermal design of the cask could meets the transport standard. The result is compared with the fire-experimentation, which shows the calculation model is conservative and rational.
Research on Method of Point Contact Modification between Spherical Fuel Particles
Guo Zixuan, Sun Zhongning, Zhang Nan
2017, 38(5): 164-168. doi: 10.13832/j.jnpe.2017.05.0164
Abstract(11) PDF(0)
Abstract:
Structured packed bed models were established for CFD simulation of convective heat transfer under turbulence condition, and the bridge method was used to deal with the problem of point contact between spherical fuel particles, to investigate the effects of cylindrical bridge size on the characteristics of flow pressure drop and heat transfer of packed bed channels. Based on the validation of experiment results, proper bridge size range was determined. The results showed that reducing the bridge size increases the calculated pressure drop gradient results in BCC and FCC packed beds; with bridge diameter not larger than 0.1 particle diameter, bridge size does not affect the average characteristics of flow pressure drop and heat transfer evidently; it is proper to choose 0.1 particle diameter as bridge diameter for flow and heat transfer simulations inside the packed bed.
Development of a 3D Multiphysics Coupled Computation Model for TRISO Fuel Particle
Chen Ping, Li Wei, Li Yuanming, Tang Changbing, Li Wenjie, Zhou Yi
2017, 38(5): 169-174. doi: 10.13832/j.jnpe.2017.05.0169
Abstract:
TRISO fuel particle is the area where fission reaction occurs for high-temperature gas-cooled reactor fuel and FCM accident-tolerant fuel pellet. To investigate the complicated behaviors of TRISO fuel particle under irradiation, a 3 D multiphysics-coupled computation model for fuel performance analysis was developed based on the COMSOL finite element software. Using the irradiation-dependent material properties and behavior models, this model is capable of simulating the complicated in-pile thermo-mechanical behavior under steady-state and accident conditions, important phenomena such as fission gas release, production of CO gas and diffusion of fission products, as well as failure probability of fuel particle. The IAEA CRP-6 benchmarks were employed to validate the model, which showed good agreement of the results between COMSOL and other codes; the simulation of TRISO fuel particle conducted by using BISON was also consistent with the results obtained by the COMSOL model, which indicated the reasonability of the model in the present study.
Study on Irradiation Behavior of Fuel Rods with FeCrAl Cladding
Gao Shixin, Li Wenjie, Chen Ping, Jiao Yongjun, Zhou Yi, HE Liang
2017, 38(5): 175-177. doi: 10.13832/j.jnpe.2017.05.0175
Abstract(16) PDF(0)
Abstract:
Due to the excellent anti-corrosion performance of FeCrAl stainless steel, it has become one of the most important and leading cladding options as accident tolerant fuel(ATF) cladding material. This paper focuses on the in-pile performance analysis of fuel rods with FeCrAl cladding, and provides suggestion on the future FeCrAl study and its application. This paper makes an introduction about FeCrAl cladding background and performance model, and uses FUPAC code to make analysis on the in-pile steady state performance. The results show that the core design criteria can be satisfied.
Study on Method of Evaluating PWR Fuel Assembly Leaf Spring Hold-down System
Pu Cengping, GenG Fei, Huang Chunlan, Pang Hua, Qi Min, PenG Yuan, ZhenG Meiyin
2017, 38(5): 178-181. doi: 10.13832/j.jnpe.2017.05.0178
Abstract:
The nonlinear phenomenon of the hold-down system is analyzed and some design response measures are proposed. Then, the method for evaluating the PWR fuel assembly leaf spring hold-down system is put forward. Last, the method is used for the design optimization of the leaf spring hold-down system of the floating NPP fuel assembly.
Design of Contact-Type Thickness Detection System for Neutron Poison Plate
Gu Mingfei, Qing Tao, Li Quan, QiN Mian, Wang Haoyu
2017, 38(5): 182-186. doi: 10.13832/j.jnpe.2017.05.0182
Abstract(19) PDF(0)
Abstract:
Spent fuel storage grid(storage grid), as one of the important equipments in the storage system of spent fuels, is placed in the spent fuel pool. To improve the storage capability per pool area for spent fuel, plate-like materials which contain some neutron poisons(such as boron aluminum and boron stainless steel) are made into quadrate storage tubes by sheet metal process and welding. The storage grid is usually made up of multiple storage tubes. The thickness of neutron poison plate which influences the safety of spent fuel storage is an important parameter of storage grid and should be measured accurately in the manufacturing process. In this paper, the contact-type measuring method is employed to design the thickness detection system for neutron poison plate by FEMA technological analysis, combining with the modern automatic detection and control technology and database technology. The designed thickness detection system, supporting one button control, can automatically finish the process of thickness detection, recording and statistic for the neutron poison plate, and improving the detecting efficiency and quality.