核动力工程

General Technology Features of Reactor Core and Safety Systems Design of HPR1000

Yu Hongxing, Zhou Jinman, Leng Guijun, Deng Jian, Liu Yu, Wu Qing, Liu Wei

2019, 40(1): 1-7. doi: 10.13832/j.jnpe.2019.01.0001

Abstract(1056)

Abstract:
HPR1000 is an advanced pressurized water reactor developed by China National Nuclear Corporation (CNNC) with fully independent intellectual property rights. In this paper, the production process of HPR1000 are firstly introduced, and then the general technology features of reactor core and safety systems design are discussed, including reactor core neutronics design, thermal-hydraulics design, safety systems design, experiment verification and the continuously design improvement and optimization of HPR1000, etc. With the brand new advanced design philosophy and technology, the economy, flexibility and safety of HPR1000 nuclear power plant regarded as a third generation nuclear power technology are enhanced.

Development and Application of a New CHF Correlation for PWR Fuel Assembly

Liu Wei, Peng Shinian, Jiang Guangming, Liu Yu

2019, 40(1): 8-11. doi: 10.13832/j.jnpe.2019.01.0008

Abstract(397)

Abstract:
Based on the bundle CHF test database of NPIC, the minimum DNBR point method is applied in this study to develop a new CHF correlation named CF-DRW correlation for the PWR fuel assembly to couple with the subchannel code CORTH. The results of typical accidents analysis indicate that the DNBR thermal margin of CF-DRW correlation is almost equivalent or even larger than the FC-2000 correlation.

Development and Verification of Multi-Group Constants Processing Module in NECP-Atlas

Xu Jialong, Zu Tiejun, Cao Liangzhi, Wu Hongchun

2019, 40(1): 12-17. doi: 10.13832/j.jnpe.2019.01.0012

Abstract(548)

Abstract:
Multi-group constants are the basis of the deterministic physics calculation, and the accuracy will directly affect the subsequent calculations. The methods of multi-group constants processing are developed rapidly, and the study on the processing methods and relevant code developing have great importance. Multi-group constants contain multi-group cross sections, transfer matrices, fission constants, etc. Based on the evaluations and the reaction rate conservation theory, the neutron slowing down equation is solved by the recursive hyper-fine group method and the nuclide mixing method is proposed. A module named Group_collapse for processing multi-group constants is developed. Numerical results show that the theories are correct, and the multi-group constants can be available for the deterministic physics codes which are based on the conventional two-step method and the one-step method.

Influence of Gas State Equations on Acceleration Parameter for Heat Transfer to Supercritical Carbon Dioxide Flowing in Heated Tubes

Liu Shenghui, Huang Yanping, Liu Guangxu, Wang Junfeng, Wang Jinyu

2019, 40(1): 18-22. doi: 10.13832/j.jnpe.2019.01.0018

Abstract(368)

Abstract:
Strong acceleration effect has an obvious influence on the convective heat transfer to supercritical carbon dioxide flowing in a heated tube. Acceleration parameter (A) is an important dimensionless number to represent the level of acceleration effect quantitatively. In the process of developing acceleration parameter, carbon dioxide state equation is needed. The difference was analyzed, caused by choosing of different state equations, the ideal state equation or van der Waals equation. The evaluation, with the help of experimental data, shows that A_van-der (based on van der Waals equation) is able to predict the local deteriorated heat transfer as a result of strong flow acceleration better than A_ideal (based on ideal state equation). Analysis showed that the volumetric expansion factor and the volumetric compression factor, with which the acceleration factor should be developed, indicate the essence of acceleration effect.

Transient Analysis on Unprotected Partial Flow Blockage of Hottest Fuel Assembly for A Small Natural Circulation Lead-Cooled Fast Reactor

Zhao Pengcheng, Liu Zijing, Yu Tao

2019, 40(1): 23-27. doi: 10.13832/j.jnpe.2019.01.0023

Abstract(290)

Abstract:
Corrosion of cladding and structural materials by liquid lead or LBE is one of the key issues restricting the development of lead-cold fast reactor. A primary cooling system analytical model of 100MW_th small modular natural circulation lead-cooled fast reactor named SNCLFR-100 is established with ATHLET code, and the unprotected transient of partial flow blockage of the hottest fuel assembly was analyzed. The results of analysis indicate that the mass flow rate of the hottest fuel assembly will drop to about 50% of the rated value when the blocking rate β reaches 0.6, while the maximum temperature of the hottest pin cladding will reach 650℃. When β reaches 0.9, the mass flow rate of the hottest fuel assembly will fall to about 12.6% of the rated value, and the maximum temperature of the hottest pin cladding will exceed the cladding material melting point 1400℃, and cladding melting will occur in the hottest fuel assembly.

Modification and Verification of Critical Flow Module for LOCA Analysis Code

Wang Jie, Liu Dong, Liu Ying, Lu Tianyu, Wu Dan

2019, 40(1): 28-32. doi: 10.13832/j.jnpe.2019.01.028

Abstract(236)

Abstract:
The conservative analysis method in the analysis of loss of coolant accident(LOCA) is not conducive to improving the economic efficiency of nuclear power plants. In order to satisfy the evaluation requirements for LOCA of nuclear power plants in 10CFR50 Annex K, the model was modified to meet the evaluation requirements for LOCA while increasing the design margin based on the best estimation code of RELAP5. Because Appendix K involves large models, this paper mainly studies the method of modifying and verifying the LOCA code model. The revision of the critical flow model of RELAP5 code was carried out. A conservative Moody two-phase critical flow model was added. At the same time, the Zaloudek model for critical flow calculation was added. The separate effect tests of Marviken and Edward’s pipe and the integral effect test of Bethsy were used to verify the code. The results show that the added model is reliable enough to simulate the critical flow phenomenon in the discharge process.

Fluidelastic Instability of Rotated Triangle Tube Bundles Subject to Two-Phase Cross-Flow

Jiang Tianze, Li Pengzhou, Ma Jianzhong

2019, 40(1): 33-36. doi: 10.13832/j.jnpe.2019.01.033

Abstract(155)

Abstract:
In order to predict more accurately the critical flow of fluidelastic instability, the motion equation of a cantilevered tube is established utilizing quasi-static fluid force coefficients of two-phase flow. After Galerkin discretization of the analytical model, the critical flow velocity of each void fraction is obtained by solving the characteristic equations. Moreover the response of cantilevered tube is solved using Runge-Kutta algorithm. Numerical results show that the critical velocity of instability increases with the increasing of the void fraction, and the numerical results agree with experimental results, which proving the analytical model utilizing quasi-static fluid force coefficients of two-phase flow is available for predicting the critical velocity of fluidelastic instability.

Research on Fuel Loading Operation Scheme for Annual Refueling Core of M310 Advanced Nuclear Power Plant

He Taifeng, Liu Xinping

2019, 40(1): 37-41. doi: 10.13832/j.jnpe.2019.01.0037

Abstract(563)

Abstract:
This paper discusses the difficulties of the loading operation process for the annual refueling core of M310 improved nuclear power unit, and studies the arrangement of the annual refueling core, the deformation form of the fuel assembly and the loading measures of the M310 improved nuclear power unit. Taking the loading operation scheme of the second overhaul core of Unit 1 of Fuqing Nuclear Power Plant as an example, based on the analysis and loading assumptions of the deformation characteristics of the fuel assemblys, the existing "snake" loading method is analyzed in depth and an optimization improvement scheme for correcting the loading order of the last three rows of fuel assemblies is proposed. After the verification, the optimized annual refueling core loading operation scheme is feasible, which can significantly improve the safety and efficiency of the loading operation.

Feasibility Study on Conceptual Design of Dual Fluid Fast Reactor

He Xun, Zeng Chang, Yu Xiaoquan, Du Zhuoqi, Rafael Macián-Juan

2019, 40(1): 42-47. doi: 10.13832/j.jnpe.2019.01.0042

Abstract(314)

Abstract:
The development of DFFR stays in the phase of conceptual design, which leads to the main discussion of this work that is to quantitatively evaluate the feasibility of this reactor type and the effects of those parameters on the reactor physical behavior. The evaluation includes two parts: the criticality calculation based on the Monte Carlo theory and the thermal-hydraulic calculation based on the heat transfer theory. The result shows that the DFFR can achieve the criticality with expected operational parameters and the feasibility of this reactor is therefore preliminarily proven.

Determination for ²³⁵U Enrichments of Same Geometry and Mass Uranium Components by BP Neural Network

Ren Lixue, Liu Zhigui, Zhou Zhiru

2019, 40(1): 48-50. doi: 10.13832/j.jnpe.2019.01.0048

Abstract(204)

Abstract:
To identify metal uranium components with same geometry and mass but different enrichments, ²⁵²Cf source noise analysis was used to gain the neutron time correlation coincidence measurement. Based on the analysis and processing of neutron time correlation coincidence measurement, the characteristic parameters are determined. Through the method of BP neural network, we determined the ²³⁵U enrichments of same geometry and mass but different enrichments uranium components. Result shows that by the method of BP neural network, we can diagnose the ²³⁵U enrichments effectively.

Research on Preparation Method for Zr-4 Tubes with Different Hydride Orientation

Chen Boquan, Peng Qian, Zeng Zihan, Zhao Suqiong, Hong Xiaofeng, Qiu Shaoyu, Xu Chunrong

2019, 40(1): 51-55. doi: 10.13832/j.jnpe.2019.01.0051

Abstract(203)

Abstract:
Zr-4 tubes with 140±20 ppm(1 ppm=1 μg/g) and 260±20 ppm hydrogen were prepared by using electrolysis hydrogenating method, and these tubes were further undergone the effect of hydride stress reorientation through loading high pressure gas. Finally, Zr-4 tubes with different hydride orientation were received. The results showed that tubes with 140±20 ppm and 260±20 ppm hydrogen can be obtained via electrolysis with 105 mA/cm²×2 h and 110 mA/cm²×(4 h+50 min) respectively. When the temperature cycle is 400-200℃ and actual heating and cooling speeds are about 10℃/min and 0.75℃/min respectively, and tubes with the hydride orientation up to 0.7 can be obtained by just one thermal cycle through adjusting the pressure and the holding time.

Study on Fabrication and Microstructural Analysis of U₃Si₂ Fuel Pellets

Zhang Xiang, Liu Guiliang, Liu Yunming, Xiao Hongxing, Liu Yu, Chen Rong, Zhang Ruiqian

2019, 40(1): 56-59. doi: 10.13832/j.jnpe.2019.01.0056

Abstract(717)

Abstract:
Uranium silicide is formed from ingot of uranium and silicon in near stoichiometric quantities by arc melting. And then uranium silicide pellets were produced by conventional powder metallurgy. The effect of pellet compression molding and sintering on density and the microstructure of U₃Si₂pellets were studied. It turned out that,0.5% of PEG was added to be binder and the green pellets was got with a pressure ranging from approximately 260 to 300 MPa, then the pellets were sintered with a temperature of 1550°C for 2-4h. To this end, high phase purity U₃Si₂with density 11.4 g/cm³,which is above 93% of theory density, is produced. The pellet size is uniform and the average grain size is about 60μm, but there is a little U or UO₂ impurity in sample U₃Si₂ pellet. The thermal conductivity of U₃Si₂ pellet is superior to UO₂, and rise approximate linearity with the increasing of temperature.

Effect of Hydride Orientation on Tensile and Bursting Properties of Zr-4 Cladding Tubes at Room Temperature

Chen Boquan, Peng Qian, Hong Xiaofeng, Liu Ranchao, Qiu Shaoyu, Xie Huaiying, Chen Le

2019, 40(1): 60-64. doi: 10.13832/j.jnpe.2019.01.0060

Abstract(496)

Abstract:
The effect of hydride orientation on mechanical properties of Zr-4 tubes with hydrogen contents with 140±20 μg/g and 260±20 μg/g was investigated by tensile and bursting tests at room temperature. The results showed that with the two studied hydrogen contents, the hydride orientation had little influence on tensile properties and bursting strengths at room temperature. The circumferential ductility was extremely sensitive to the hydride orientation. The bursting elongation at room temperature was degraded remarkably with the increasing of the hydride orientation factor. The bursting elongation at room temperature for tubes with 260±20μg/g hydrogen decreased faster compared with that for 140±20 μg/g.

Tritium Permeation in RAFM Steel

Lu Guangda, Xiang Xin, Bao Jingchun, Fan Dongjun, Zhang Guikai, Chen Changan, Tang Tao

2019, 40(1): 65-68. doi: 10.13832/j.jnpe.2019.01.0065

Abstract(205)

Abstract:
The tritium permeability of CLAM steel, one of the low activated ferrite/martensite steels (RAFM steels) made in China, have been experimentally measured by means of the gas evolution permeation technique over the temperature range of 573-823 K. The resultant permeability F_T is 2.57×10^-8exp(-38639/RT), the deduced diffusivity D_T is 1.17×10^-7 exp(-22011/RT) and Sieverts’ constant S_T is 2.2×10^-1exp(-38639/RT). Again, there is an obvious isotope effect on D-T mixture permeation in which the permeability of deuterium is larger than that of tritium. The permeation separation coefficient α_DT is 1.42 for D-T, and α_HTis3.76 for H-T.

Research on UF₄ Producing Techniques by Fluoridation of U₃Si₂ Powder

Zhang Fan, Yu Xiaochuang, Li Tao, Guo Bolong

2019, 40(1): 69-73. doi: 10.13832/j.jnpe.2019.01.0069

Abstract(239)

Abstract:
There are many kinds of unqualified uranium containing materials with large stock. In order to improve the utilization rate of uranium and meet the stable supply of materials needed for fuel element production, uranium recycling is needed. It is introduced that U₃Si₂ powder can be converted into uranium oxides in calcinations process, after which the uranium oxides will react with solid NH₄F to generate UF₄. There are several factors in the producing process such as the amount of fluoride, temperature, reacting time, etc. Thus, the optimal reacting conditions are found. According to all these experiments, it can be inferred that U₃Si₂ powder after calcination and oxidation would react with solid NH₄F or NF₄HF₂ to generate UF₄. The UF₄ products will be up to the standard, meanwhile the amount of UO₂F₂ in UF₄ can be reduced to lowest level if the reacting temperature remained at 500℃ and preserving for 4.5 hours.

Study on Process and Properties of Pulse Laser Prepared Cr Coating for Accident Tolerant Fuel Claddings

Li Rui, Liu Tong

2019, 40(1): 74-77. doi: 10.13832/j.jnpe.2019.01.0074

Abstract(418)

Abstract:
This paper introduces the latest ATF cladding achievements of China General Nuclear Power Corporation (CGN). The Cr coating of Zircon with different thickness was prepared under different power by pulse laser cladding technique. The results obtained so far show that Cr coating can achieve good high temperature oxidation resistance which prepared using pulse laser cladding technology under the parameters such as pulse laser power for 50~60 W, pitch for 0.8~0.9 mm and angular velocity for 10 mm. The mechanism analysis shows that the Cr coating has good compatibility with the Zirconium alloy at 1200℃, and the protective effect of Cr coating is directly controlled by the coating quality.

Design and Numerical Research of Thermal Insulation Water Layer for Built-in Pressurizer

Zeng Chang, Sui Haiming, Ren Yun, Zhong Fajie

2019, 40(1): 78-81. doi: 10.13832/j.jnpe.2019.01.0078

Abstract(256)

Abstract:
A thermal insulation water layer is designed for the built-in pressurizer of ACP100+ small model reactor. The flow and heat transfer characteristics of heat insulation water layer are researched using numerical method, and the temperature and velocity distributing during power operation steady condition and during the drop power operation transient condition are analyzed. The results show that, the inner fluid is with weak flow and thermal conductive ability, and the structure of thermal insulation water layer isolate the high temperature from the low temperature fluid effectively.

Numerical Analysis of Dynamic Response for SG LOCA Shaking

Huang Qian, Yu Xiaofei, Qi Huanhuan, Feng Zhipeng, Jiang Naibin, Song Haiyang

2019, 40(1): 82-86. doi: 10.13832/j.jnpe.2019.01.0082

Abstract(197)

Abstract:
Through reasonable simplification and equivalence finish, a detailed nonlinear finite element model of steam generator (SG) of a domestic 3rd generation nuclear power plant (NPP) is established. This model is then connected with the reactor coolant loop (RCL) to carry out the analysis of dynamic response for SG LOCA shaking. By calculation, the maximum absolute stresses of SG heat transfer tube bundles and its variation with tube diameter and reacting forces of upper supports are obtained. In order to study the effect of SG decoupling from the RCL on the shaking dynamic response, a comparative study of decoupling /coupling methods is developed. Results show that SG decoupling has a significant impact on the calculation result, and the calculation method of coupling is more closer to the real situation and should be recommended.

Research of Operating Behavior and Response Time in Digital Control Room of Nuclear Power Plants

Li Linfeng, Xiang Xiaohan

2019, 40(1): 87-90. doi: 10.13832/j.jnpe.2019.01.0087

Abstract(620)

Abstract:
In this paper, the operator simulator retraining in a certain nuclear power plant with a digital control room is to be carried out a field observation, in order to collect image data of operating behavior and response time in the digital operating platform. According to the applied behavior analysis technology of image data, the various operating behaviors are grouped to obtain the mean of response time. The operation mode for digital operating platform and the response time of each operating action group are finally determined. The research results provide a strong theoretical and practical basis for defining the qualitative description of human reliability analysis and reducing the uncertainty of quantitative analysis.

Analysis of Emergency Crew Behavior in the Process of Severe Accident Mitigation in Nuclear Power Plants

Chen Shuai, Zhang Li, Qing Tao, Li Linfeng, Liu Zhaopeng, Niu Maolong

2019, 40(1): 91-96. doi: 10.13832/j.jnpe.2019.01.0091

Abstract(267)

Abstract:
In order to analyze the behavior characteristics of the nuclear power plant crew when dealing with severe accidents, focusing on the study for the particularity of Severe Accident Management Guideline, combined with field investigation and interviews with operator and emergency technical support personnel, this paper establishes a crew decision-making model in the process of severe accident mitigation, and identifies the key influence factors of decision makers and executors, which laid a firm foundation for the studies of human reliability analysis method in severe accidents.

Analysis of R Bank Frequent Action Problem in Yangjiang Nuclear Power Plant

Xiang Hongyi, Liu yang, Huang Zhenguang, Chen Jiancai, Chen Kailin

2019, 40(1): 97-100. doi: 10.13832/j.jnpe.2019.01.0097

Abstract(195)

Abstract:
In the 2^nd and 3rd nuclear fuel cycles of Unit 1,Unit 2 and Unit 3 in Yangjiang Nuclear Power Plant, , R bank acts frequently, seriously affecting the safety with operation of the unit. In this paper, the logic control principle of R bank and the loading of core are analyzed, and the transfer function of first-order inertia, lead lag and differential link is converted into difference equation. The theoretical calculation is carried out before the parameter is modified, and the temperature ±4℃ disturbance simulation test upon 87%FP power platform is carried out on the plant simulator. The test results show that the modified parameters are correct and can solve the frequent action problem of R bank. After the modification of the parameters for Unit 2, the R bank acts normally, and R bank action frequency is about 1 time per 2 days.

Research and Application of the Blowdown Cave Hydraulic Flushing Device of the Horizontal Steam Generator

Yan Weifeng, Liu Jianmin

2019, 40(1): 101-104. doi: 10.13832/j.jnpe.2019.01.0101

Abstract(181)

Abstract:
In order to clean up corrosion deposits in the blowdown cave of secondary side of steam generators in Unit 1&2 of Tianwan nuclear power station and reduce its corrosion risk to base metal of shell and weld joints, a hydraulic flushing device was developed and applied during the outages. Through hydraulic flushing, the corrosion products accumulated in the blowdown cave of steam generators for more than ten years were cleared. The inside walls of the blowdown cave and the weld joints were checked and they are in good condition, which improved the operational safety and reliability of the steam generators.

Analysis and Countermeasures of Vibration for Variable Frequency Pump in Nuclear Power Plants

Dong Baoze

2019, 40(1): 105-109. doi: 10.13832/j.jnpe.2019.01.0105

Abstract(186)

Abstract:
During the startup process of a nuclear power plant, a variable frequency pump has a large vibration fault in a frequency range. The vibration is not conducive to the long-term operation of the pump, so a knocking test is carried out to get the natural frequency of the pump. It is determined by the vibration spectrum analysis that the pump occurs a structural resonance. Based on the theory of structural resonance, the method of changing the stiffness or mass of the motor support and calculating the natural frequency after the finite element analysis is given. Through this method, the vibration of the variable frequency pump after the treatment of the vibration fault is obviously smaller, and it can meet the requirements. This study provides an important reference for the fault diagnosis of rotating equipment.

Research and Design of Rod Control Cabinet in Nuclear Power Plants Based on Digital Control

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

2019, 40(1): 110-115. doi: 10.13832/j.jnpe.2019.01.0110

Abstract(300)

Abstract:
There are disadvantages such as poor applicability, slow dynamic response of load output and poor anti-interference ability of the rod control power supply cabinet in operation. This paper redesigns a new type of power cabinet based on programmable logic controller (PLC) and digital signal processing (DSP) technology which adopts fast energy absorption loop and double closed-loop control method in the main circuit. Then the coordination test in the cold and hot states between the designed cabinet and ACP1000 control rod drive mechanism(CRDM) is carried out. The test result proves that the function of the cabinet satisfies the application requirements of generation-III nuclear plants and effectively solves the disadvantages of the existing cabinet.

Research on Calculations of Radioactive Source Terms for PWR Nuclear Power Plants

Xu Yanfeng, Zhang Pengfei

2019, 40(1): 116-119. doi: 10.13832/j.jnpe.2019.01.0116

Abstract(623)

Abstract:
In this paper, the method to calculate the radioactive source terms released to the environment is provided based on the analysis results of samples in the containment atmosphere during an accident at a PWR Nuclear Power Plant, relevant factors are discussed, and the validity is verified through the comparison of results calculated by this method and the code RASCAL4.2, which is a radiological assessment system for consequence analysis. Meanwhile, some deficiencies of the code RASCAL4.2 are founded during the verification, and improvement suggestions are provided for the code.

Numerical Research on Source Inversion for Nuclear Accidents Based on Variational Data Assimilation Coping with Truncated Total Least Squares

Liu Yun, Liu Xinjian, Li Hong, Fang Sheng, Mao Yawei, Qu Jingyuan

2019, 40(1): 120-125. doi: 10.13832/j.jnpe.2019.01.0120

Abstract(278)

Abstract:
In a nuclear accident, release source term is the key issue of nuclear emergency and consequences assessment. Source inversion for nuclear accidents is a way to estimate the source term by using the environmental monitoring data during the accident. Since it does not rely on the reactor state parameters, it receives a wide attention after the Fukushima nuclear accident. The modeling of source inversion for nuclear accidents based on variational data assimilation (VAR) is able to get the global optimal result. However, its result is great influenced by the error of the atmospheric dispersion model used in source inversion. In order to reduce the effect of the dispersion model error on the estimated source term, VAR coping with truncated total least squares (TTLS-VAR) was proposed. It is able to correct the dispersion model operator and monitoring data vector to enhance the accuracy of source inversion. The TTLS-VAR was verified using the date of wind tunnel experiment. The results demonstrated that the accuracy of the result calculated by the TTLS-VAR is higher than that by the VAR.

Study on Impact of Leakage of Toxic and Hazardous Gases on Habitability of Main Control Room Based on ALOHA

Li Hui, Li Chaofeng, Shi Yanming, Xiong Min

2019, 40(1): 126-130. doi: 10.13832/j.jnpe.2019.01.0126

Abstract(178)

Abstract:
Based on a nuclear power plant in China and the assess principle of the Regulatory Guide 1.78-Evaluating the Habitability of a Nuclear Power Plant Control Room during a Postulated Hazardous Chemical Release, the U.S. nuclear safety guide, we collected and screened the chemicals used in this project, and use ALOHA to calculate the concentration of toxic and hazardous gases entering the main control room when a leak occurs. Finally, the impact of the leakage on the habitability of the main control room is evaluated. The simulation results show that, since the nuclear island is enclosed and the air vents of the main control room is inside of the nuclear island, when a toxic and hazardous gases leak occurs, the concentration of toxic and hazardous gases near the air vents is lower than the limits, and will not have a significant impact on the habitability of the main control room.

Study on Dynamic Adsorption of Gaseous Iodine by Silver Loaded Mordenite and Alumina

Xiong Wei, Cao Qi, Wang Haijun, Chen Yunming, Wu Wangsuo, Zhang Jinsong

2019, 40(1): 131-134. doi: 10.13832/j.jnpe.2019.01.0131

Abstract(227)

Abstract:
A dynamic adsorption device was designed for high humidity and high nitrogen oxide iodine gases produced by nuclear facilities. Silver-loaded mordenite and silver-loaded alumina were prepared for adsorbing gaseous iodine. The adsorption properties of silver-loaded mordenite and silver-loaded alumina for gaseous iodine were studied by changing the amount of silver and the concentration of iodine vapor. The experimental results show that the saturated adsorption capacity of iodine by mordenite with silver content of 15.2% is 178.4 mg/g under the conditions of 30℃, 100% relative humidity and 0.1% NO₂. The adsorption effect of silver-loaded mordenite on gaseous iodine is better than that of silver-loaded alumina. It is a complex process mainly based on chemical adsorption and physical adsorption.

Development of Nuclear Reactor Design Software Verification Database System

Liu Ying, Feng Bo, Cao Guohai, Tang Lei, Feng Jintao, Lu Jiachuan, Yu Yang, Zhou Yueshan, Qiang Shenglong

2019, 40(1): 135-139. doi: 10.13832/j.jnpe.2019.01.0135

Abstract(123)

Abstract:
The nuclear reactor design software verification database system was developed in this paper, aiming at the actual demands of professional data during software verification from many majors, such as reactor core physics, thermal-hydraulic analysis, fuel design, etc. Logic structure and technical framework of the system are presented. The unified data modeling way is adopted to analyze, design and realize the nuclear reactor design software verification database system from three management levels of data, system, and safety management. The developed system solved the drawbacks and deficiencies of the verification data that lack of organization standardization, data repeat-ability, and data low-usage. It provides the vital technical support to finish subsequent jobs of software verification.

Analysis of Background Noise Power Spectrum for Steam Generator of Sodium Cooled Fast Reactor

Cao Yunqi, Liu Guijuan, Liu Zhiguo

2019, 40(1): 140-143. doi: 10.13832/j.jnpe.2019.01.0140

Abstract(189)

Abstract:
In order to understand the variation of fluid operation and internal composition in the steam generator of sodium-cooled fast reactor, the AR model of modern spectrum estimation and the periodic graph method of classical spectrum estimation, Welch improved periodic graph method and BT method are introduced to compare and analyze the original background noise data produced by the operation of sodium cooled fast reactor steam generator. The results show that AR model is more effective in estimating the background noise of sodium-water reaction.

Methods to Determine Intensity of Historical Earthquake in Nuclear Power Plant Sites

Wang Ji, Zhang Yushan, Yan Jingru, Zhang Yujie

2019, 40(1): 144-146. doi: 10.13832/j.jnpe.2019.01.0144

Abstract(399)

Abstract:
The three methods for determining the influence intensity of the historical earthquake in nuclear power plant sites are the method for extrapolation of the seismic isosematics, the methods of judging data of earthquake damage investigation in ancient buildings, and the method of investigation and judgment of earthquake damage near the sites. Taking a nuclear power plant site in Hebei Province as an example, the influence intensity of the 1679 Sanhe and Pinggu M8 earthquake on the site was judged by the method of historical earthquake and isoseismal inference, which was Ⅶ degree, the influence intensity of the site was assessed by the investigation data of ancient buildings damaged by Tangshan M7.8 earthquake in 1976 as VI degree, and the influence intensity of the site was assessed by household-to-household earthquake damage investigation as V degree.

Study on Transient Performance in Loss of Coolant Accident for AP1000 SFP

Duan Yongqiang, He Xun, Jing Futing, Cai Zhiyun, Yu Xiaoquan

2019, 40(1): 147-151. doi: 10.13832/j.jnpe.2019.01.0147

Abstract(243)

Abstract:
The coupled mathematical thermal equilibrium model for main equipment in the spent fuel pool (SFP) is established and the transient change of the temperature of SFP water in various loss of coolant accidents is studied based on AP1000 power plant. The results show that the fuel rods are exposed after 24h when the loss of coolant accident occurs during the unloading of the whole reactor, while the fuel rods are exposed after 213h when the loss of coolant accident occurs after the fuel loading operation. The results give a reference response time for the operator to take countermeasures for the loss of coolant accident for the spent fuel pool.

Comparative Study of CHF Correlations in Upflow Boiling Vertical Round Tube under High Pressure

Liu Wei, Peng Shinian, Jiang Guangming, Liu Yu, Shen Caifen, Shan Jianqiang

2019, 40(1): 152-155. doi: 10.13832/j.jnpe.2019.01.0152

Abstract(229)

Abstract:
Based on the experiment database of upflow boiling in the vertical round tube from 15 MPa to critical pressure of water, the Katto, Bowring, Hall-Mudawar, Alekseev correlation and LUT-2006 are comparatively studied. With the error analysis of the CHF correlations to the experiment database, the applicability of these correlations is estimated and the parametric trends of CHF with pressure from 15 MPa to critical pressure are proposed. Findings of this study have a positive effect on further development of CHF prediction method, especially under high and critical pressure area.

Research on Failure Mechanism of M3 Fuel under Steady-State Operation Conditions

Li Yuanming, Tang Changbing, Yu Hongxing, Xin Yong, Chen Ping, Zhou Yi

2019, 40(1): 156-161. doi: 10.13832/j.jnpe.2019.01.0156

Abstract(165)

Abstract:
In order to optimize the design of M3 fuel and further improve its reliability in LWR environment, it is necessary to research its failure mechanism under steady-state operation conditions. Based on the ABAQUS software, a 3D numerical simulation analysis method for the irradiation-thermal-mechanical coupling behavior of M3 fuel was established by the way of secondary development. With the help of this established analytical method, the failure mechanism of M3 fuel under steady-state operation conditions was carried out. According to the research results, during the steady-state operation conditions, the failure mechanism of M3 fuel is mainly dominated by the failure of IPyC layer in the initial irradiation and the failure of SiC layer caused by the contact of Buffer layer and IPyC layer. This research results could provide guidance for the optimization design of M3 fuel.

Study on Characteristics of Core Barrel Shell Mode Vibration Based on Neutron Noise Analysis

Li Yun, Liu Caixue, Luo Ting, Yang Taibo

2019, 40(1): 162-166. doi: 10.13832/j.jnpe.2019.01.0162

Abstract(157)

Abstract:
The vibration of the reactor core barrel reflects the vibration condition and equipment stability of the core barrel and corresponding reactor internals. This paper mainly focuses on the vibration characteristics of the core barrel shell mode with neutron noise technique and time-frequency signal analysis, and the change trend of shell mode characteristic quantities are obtained. The result shows that the vibration frequency of the core barrel shell mode has a decreasing trend within each fuel cycle and the vibration frequency would be recovered to the initial vibration frequency of the beginning of the last fuel cycle. The research of this paper contributes to further understanding of the vibration characteristic quantities of the core barrel shell mode and its causes of formation in multi-fuel cycles. The study results lay the foundation for early fault diagnosis for the core barrel.

Analysis for Abnormal Shaft Vibration of Main Coolant Pump in Nuclear Power Plants

Li Zhen, Yuan Shaobo

2019, 40(1): 167-171. doi: 10.13832/j.jnpe.2019.01.0167

Abstract(248)

Abstract:
To solve the shaft vibration alarm problem of the main coolant pump in a nuclear power plant, the reason diagnosis and field verification are carried out. The methods of comparison analysis, spectrum analysis and orbit analysis have been applied to investigate the abnormal shaft vibration. The analysis result shows that the vibration of the pump shaft is greater than that of the motor shaft, and the vibration at the measuring points on the same horizontal plane in two different directions is similar. The abnormal vibration at Y direction of the motor shaft is caused by the failure of the cable shielding layer. It is suggested to use the soft materials, such as hemp rope for bandage in the vibration sensor cable installation. The abnormal vibration of the pump shaft is caused by large eddy motion and higher amplitude of running frequency components. In the case of the obvious vortex motion of the pump shaft, the shaft seal is suggested to be improved to reduce the vibration of the pump shaft.

Research on Current Monitoring and Fault Diagnosis Technology for Control Rod Drive Mechanism

Zeng Jie, Peng Cuiyun, He Pan, Liu Caixue

2019, 40(1): 172-175. doi: 10.13832/j.jnpe.2019.01.0172

Abstract(370)

Abstract:
The control rod drive mechanism (CRDM) is the actuator of the reactor control and nuclear safety protection system. An important measure to prevent the occurrence of control rod stuck, sliding rod and driving failure in the reactor is to effectively monitor the operation status of the CRDM. Based on the analysis of the action principle of the control rod driving mechanism, this paper realizes the current monitoring and fault diagnosis of the control rod driving mechanism by studying the detection method, signal analysis and processing method and fault identification method of the driving mechanism, which lays a technical foundation for the application of the current monitoring and fault diagnosis system of the driving mechanism.

2019 Vol. 40, No. 1