核动力工程

Study on Dynamic Characteristics of Xi’an Pulsed Reactor under Non-Pulse Transient Condition

Zhang Liang, Yuan Jianxin, Zhao Wei, Wang Baosheng, Zhang Qiang, Zhu Guangning, Jiang Xinbiao, Chen Lixin

2020, 41(5): 1-7.

Abstract(453) PDF(369)

Abstract:
The instrument and control system of Xi’an Pulsed Reactor is under the digitized reconstruction, and the code for analyzing the dynamic characteristics of non-pulse transient conditions is essential to provide the real-time variety parameters such as power and fuel temperature. Based on the previous classical analysis code of UHZr Pulsed Reactor, the model for analyzing the dynamic characteristics is built by optimizing the reactor physical model of the previous code and adding new models, and the Xi’an Pulsed Reactor Dynamic Characteristic Analysis Code (XPRDCA) is developed for non-pulse transient analysis. The experiment on the Xi’an Pulsed Reactor is performed. The calculation results of the code and the data obtained from the experiment are compared, and the effect of the temperature reactivity coefficient and the gap heat transfer coefficient on the dynamic characteristics is studied. The calculation results are in good agreement with the experiment data, and the computation speed of the code with optimized model is remarkably improved, and the code could be applied to the design and debugging of the digital instrument and control system.

Correction of CPPF Factor in QINSHAN CANDU Reactor

Wang Jun

2020, 41(5): 8-11.

Abstract(253)

Abstract:
With the aging of the unit, the core fuel cooling performance of QINSHAN CANDU reactor decreases. To ensure the safe operation of the reactor, it needs to do regular analysis of ROP TSP. The analysis result is applied by correcting the CPPF which is used to calibrate ROP detectors to continuously track thermal hydraulic, pressure tube creep and other parameters. This paper focuses on the analysis of the ROP related parameters correction method for CPPF. Based on the actual operation data of unit 1, it analyzes the comparison of relevant parameters before and after correction, as well as the influence of operation such as refueling. Finally, it summarizes the operational optimization and countermeasure that can be taken.

A Brief Method for Derivation of Streaming Term Expression for Neutron Transport Equation in Spherical Geometry Coordinate System

Yang Benlin, Chen Shi, Wan Like, Wang Ziguan, Zhao Xinyue, Hu Longxiang, Yang Song

2020, 41(5): 12-14.

Abstract(136)

Abstract:
An ingenious method has been discussed to provide a brief way for the derivation of the streaming term expression in the spherical geometry coordinate system. Instead of finding the differential relationship between the dihedral angle and the neutron transport distance, finding the geometry relationship between the dihedral angle and other plane angles is much easier. And then the differential relationship (easy to be derived) between the other plane angles and the neutron transport distance can be used to derive the expression of the streaming term in the spherical geometry coordinate system indirectly but easily. This method makes the derivation process easy and intuitive, with clear physical picture.

Development and Validation of Non LOCA Thermal Hydraulics and Three Dimensional Neutronics Coupling Code GINKGO/COCO

He Qingyun, Luo Jingyi, Chen Jun, Ren Zhihao, Peng Sitao, Zhou Zhou, Shan Jianqiang

2020, 41(5): 15-19.

Abstract(208)

Abstract:
For the needs of more detailed and accurate core modeling and thermal hydraulic analysis, based on the Non LOCA thermal hydraulic analysis code GINKGO developed on self-reliance and three-dimensional core physical code COCO, GINKGO/COCO coupling code has been developed by using Dynamic Link Library (DLL) method. This paper introduces the development principle and implementation of the coupling code, and the international benchmark for main steam line break accident (MSLB) completed by Organization for Economic Co-operation and Development (OECD) is chosen to validate the coupling code. The results show that the calculation results of GINKGO/COCO are in good agreement with the benchmark values. It is indicated that the GINKGO/COCO coupling code is with good computing performance and reliability.

Three-Dimensional Flow Field Calculation of Pressure Vessel in Small Pressurized Water Reactor

Wang Kun, Dong Xiucheng, Liu Haipeng, Zhang Xin, Yuan Jiangtao

2020, 41(5): 20-23.

Abstract(280)

Abstract:
In the reactor safety analysis process, it is important to obtain an accurate flow field inside the pressure vessel. Taking the small pressurized water reactor as the research object, the computational fluid dynamics (CFD) method was used to calculate and analyze the internal flow field of the reactor pressure vessel, and the fuel assembly flow distribution and the lower head mixing characteristics were obtained. The results show that the maximum flow distribution coefficient of the fuel assembly is 1.032, the minimum value is 0.934, and the overall flow distribution is characterized by “large in the middle and small in the edge” under the high-speed symmetrical inlet condition of the two pumps. The flow vortex of the lower head is enhanced, and the uneven distribution of the flow distribution of the fuel assembly is increased, under the high-speed asymmetric inlet condition of the pump. The minimum mixing factor of the coolant flow at the core inlet was calculated to be 0.022 due to the insufficient mixing characteristics of the lower head.

Analysis of Coupled Flow and Heat Transfer in Primary and Secondary Sides of Helical Coil Once-Through Tube Steam Generator

Liu Fayu, Zhang Xiaoying, Chen Jiayue, Chen Hu, ong

2020, 41(5): 24-29.

Abstract(363)

Abstract:
To study the flow and heat transfer characteristics of the primary and secondary sides of the helical coil once-through tube steam generator (HCOTSG) under steady state conditions, taking HCOTSG of International Reactor Innovative and Secure (IRIS) as the research object, a primary and secondary sides heat balance calculation model for steady state operation of HCOTSG is established. The influence of different secondary side feed water flow rate on HCOTSG thermal and hydraulic parameters under steady-state condition is analyzed, and the detailed thermal and hydraulic parameters in the helical tube under steady-state condition are calculated by combining the coupled thermal analysis model with the three-dimensional flow and heat exchange calculation of CFX. The relevant thermal and hydraulic parameters along the tube side of HCOTSG during steady-state operation are calculated by the thermal analysis model. The CFX simulation results show that the velocity and temperature distribution of the fluid in the cross section of the helical tube are not uniform. The temperature of the fluid inside the helix is higher than that outside the helix. The velocity of the fluid inside the helix is lower than that outside the helix. The boiling of the fluid inside the helix occurs earlier than that outside the helix. Therefore, this study has a guiding role in the accident analysis for HCOTSG steady-state operation and spiral heat exchange tube.

Experimental Study on Enhancement of Pool Boiling Critical Heat Flux on SA508 Carbon Steel with Cold Spray Coating

Qin Fei, Liu Hanzhou, Hu Lian, Chen Deqi, Zhong Dawen

2020, 41(5): 30-34.

Abstract(160)

Abstract:
In this study, a coating technique known as “cold spray” was developed to form a micro-porous coating on the SA508 Gr3 carbon steel heater, which was the prototypical material for the reactor vessel. On the rotatable experimental bench, downward facing pool boiling CHF experiments using a plate test section under atmospheric pressure was performed to investigate the quenching curves on the CHF limits with and without cold-spray coatings at different inclination angles. Quantitative measurements showed that the local CHF values for the coated surface were consistently higher than the corresponding CHF values for the bare surface, and that the use of cold-spray coatings could enhance the local CHF limits for downward facing boiling by more than 25%. In addition, the micro-porous coatings were stable enough after many cycles of heating and cooling.

Research on Method of Generating Monte Carlo Multigroup Library for Fast Reactor

Zhu Shuaitao, Ma Xubo, Xu Qian, Cao Bo, Chen Yixue

2020, 41(5): 35-39.

Abstract(161)

Abstract:
Based on the discrete angle method, a Monte Carlo multi-group cross section generation program MGXSMC was developed. This program can read the cross section data from an input file or read the cross section from a library in a specified format to generate the multi-group cross section for MCNP or RMC. The corresponding index file list can be automatically generated. The two-dimensional two-group IAEA pressurized water reactor (PWR) benchmark and lead-based fast reactor (RBEC-M) benchmark were used to verify the cross section library generated by the MGXSMC program. The calculation results show that the difference between the calculated result of the P5 order approximate multigroup section and the continuous point cross section is 24 pcm (1pcm = 10-5), and the difference of the keff result calculated by the P0 order approximate multigroup section and the continuous point section is large. This shows that the method and the program developed for the Monte Carlo Group Section Library are correct. At the same time, the neutron anisotropic scattering has a large impact on the calculation results of the lead-based fast reactor. Therefore, when the Monte Carlo Group Section library is produced, the neutron scattering angle data should be added.

Study on Thermal Conductivity Model of Dispersion Fuel

Ren Qisen, Liao Yehong, Chen Mengteng, Zhang Yongdong, Xie Yiran, Liu Tong, Liu Weiqiang

2020, 41(5): 40-43.

Abstract(266)

Abstract:
The effective thermal conductivity (ETC) of dispersion fuels plays an important role in nuclear reactor safety analysis and fuel performance evaluation. In this study, based on the theory of porous body, considering the relativity of dispersion particle distributions, an ETC model of dispersion fuels was proposed and validated. The effect of porosity, fuel volume fraction and fuel-matrix thermal conductivity ratio on the ETC were investigated. The results show that the ETC decreases along with the increasing of fuel volume fraction and porosity; the higher the fuel-matrix thermal conductivity ratio is, the less the effect of fuel volume fraction on the ETC of dispersion fuel is.

Design and Verification of Relaxation Test Device for Hold-down Leaf Spring

Luo Wenguang, Wang Yajun, Wang Wanjin, Wu Rui, Zhang Xianmeng

2020, 41(5): 44-48.

Abstract(180)

Abstract:
A relaxation test device for the hold-down leaf spring was designed considering the characteristics of the hold-down leaf spring for AFA3G fuel assembly. The reliability of the device is verified by a series of simulation tests. Finally, four groups of different recycled AFA3G fuel assemblies were selected, and the compression force and the deformation of the four compressed plate springs were accurately measured at any position in the whole stroke, using this relaxation test device. The results showed that the repetitive accuracy and the comprehensive accuracy of the device was 0.49% and 1.7%，satisfying the requirement of relaxation test.

Prediction Method for Long-Term Smt of 316 Stainless Steel

Li Changxiang, Mo Jintao, Duan Chunhui

2020, 41(5): 49-52.

Abstract(245)

Abstract:
General primary membrane stress (S_mt ) of material is the important parameter used for the mechanical analysis of high temperature reactor design, and the data of Smt at 300000 hours in the ASME code and RCC-MR code cannot meet the needs of long-life nuclear reactor design. Based on the data of allowable stress intensity (S_t), expected minimum stress-to-rupture (S_r ) and stress rupture factors (R ) at 300000 hours in the ASME code, the Smt at 500000 hours of 316 stainless steel base metal and weld required by long-life reactor design are obtained successfully by Larson-Miller extrapolation model.

Creep Crack Growth Behavior of N18 Zircaloy Thin-Walled Tubes with Double Edged Axial-Notched Cracks

Li Zhihao, Bao Chen, Wang Bo, Liu Xiaokun

2020, 41(5): 53-59.

Abstract(163) PDF(116)

Abstract:
Through the design of double edged axial-notched tube (DEAT) specimen and fixture, the expression of C* integral of DEAT specimen was obtained based on the principles of energy equivalent and load separation. A test method for the creep crack growth rate of thin-walled tubes with axial cracks has been established. The creep crack growth tests of N18 zircaloy thin-walled tubes at different load levels at 350℃ are carried out by using DEAT specimens. The results show that the creep load will significantly affect the creep crack growth rate of N18 zircaloy; the creep crack growth can be divided into two stages: steady state expansion and rapid expansion; the creep crack growth rate (da/dt) and C* integral have a good power-law relationship, and it can be used to predict the creep crack growth behavior of N18 zircaloy.

Analysis of Fatigue Time Limit Aging of Reactor Pressure Vessels

Shao Xuejiao, Xie Hai, Zhang Liping, YangYu, Du Juan, Tian Jun, Kuang Linyuan, Gao Shiqing

2020, 41(5): 60-64.

Abstract(294)

Abstract:
Based on two methods of evaluating the influence of the coolant environment on the fatigue life of equipment proposed by U.S.NRC in the management guideline RG1.207, the effects on different environmental fatigue correction factor（F_en） expressions and boundary conditions were compared between the NUREG/CR-6909 of USA and JNES of JAPAN. The difference of the environmental fatigue life assessment between environmental fatigue correction factor and environmental fatigue curve was also analyzed. Finally, the three methods were adopted in the analysis of the reactor pressure vessel inlet nozzle fatigue assessment. The methods are Fen method considering strain rate history, environmental fatigue curve method and the Fen method using conservative parameters. The results show that, compared with other two methods, the Fen method considering strain rate history can evaluate the environmental fatigue life of structures with higher accuracy.

Study on Nonlinear Beam Model of PWR Fuel Assembly

Gu Chenglong, Yang Yuying, Guo Yan

2020, 41(5): 65-69.

Abstract(180)

Abstract:
In order to describe the nonlinear characteristics of the fuel assembly, a lateral beam model of assembly is built by the finite element method. The beam model is embedded with the hysteretic model to simulate the nonlinear effect occurred at lateral deformation. The calculation shows that the bending deformation and force can be derived from the nonlinear beam model, which is beneficial for the analysis of fuel assembly accidents.

Effect of Main Bolt Break on Seal Performance, Stress and Fatigue of Reactor Pressure Vessel

Zheng Liangang, Bai Xiaoming, Shi Kaikai, Du Juan

2020, 41(5): 70-73.

Abstract(144)

Abstract:
Base on the mechanics theory and numerical simulation technique, a method to analyze the effect of the main bolt break on the stress, fatigue and seal is studied in this paper, and is adopted to evaluate and analyze the fracture influence of main bolt. The results show that this method is applicable for the analysis of the RPV safety performance induced by one bolt break or several bolts break accident, and for the determination if the nuclear reactor can be operated when similar problems occur.

Study of Application of Level 2 Probabilistic Safety Analysis in Severe Accident Management

Zhang Jiajia, Ni Man, Xiao Jun, Gong Yu, Qian Hongtao

2020, 41(5): 74-78.

Abstract(139)

Abstract:
Level 2 Probabilistic Safety Analysis (PSA) can be used to quantitatively assess the risk of severe accident and is a good tool to evaluate the severe accident management. By studying the general method and procedure for the application of level 2 PSA in severe accident management, taking an improved generation-Ⅱnuclear power plant as an example, the “primary loop depressurization operation ” and the “ primary loop emergency water injection” in severe accident management guideline are quantitatively evaluated. Analysis shows that performing the “primary loop depressurization operation” immediately after entering the severe accident management guideline can greatly reduce the risk of large radioactive release, and performing “primary loop emergency water injection operation” contributes greatly to reducing the risk of large radioactive release in the slower accident sequence. The study shows that there still has further improvement room in severe accidents management for nuclear power plants in China.

A Brief Introduction of Design Optimization for AP1000 Condensation Return

Ma Baisong, Zhuang Yaping, Qie Weiqing

2020, 41(5): 79-83.

Abstract(162)

Abstract:
During any non-LOCA event, the average temperature of the reactor cooling system (RCS) should be decreased to 215.6℃within 36 hours in AP1000 nuclear power plant. However, this goal cannot be achieved because the rate of condensate return was far lower than that expected. Through the analysis and verification of the dome condensation drop test, the path of condensation loss was determined. Therefore, a series of the design modifications to the polar crane girder, the internal stiffener and the condensation return gutter were carried out. Lastly, the safety shutdown temperature evaluation showed that the AP1000 plant could reduce the RCS average temperature to 215.6℃ within 34.6 hours after the loss of normal feedwater and the loss of off-site power concurrently.

Simulated Analysis of Pressure Control and Overpressure Problem of Reactor Cold Start-up with Nuclear Heating

Qing Xianguo, Xiao Kai, Huang Ke, Chen Guanyu, Li Yiliang, Chen Zhi

2020, 41(5): 84-88.

Abstract(140)

Abstract:
Based on the control requirements of the reactor cold start-up process with nuclear heating, the automatic pressure control of reactor cold start-up with nuclear heating is studied in this paper, and the method for the system pressure automatic control in the process of cold start-up with nuclear heating are studied, and the automatic control method is designed and the control simulation verification is completed. At the same time, the problem of overpressure in cold start-up with water compaction state is simulated and analyzed. The interlock control methods to prevent overpressure accidents are proposed. The results show that, when the reactor power does not exceed a certain power level, the automatic pressure control method can realize the effective control of the reactor pressure during the cold start-up with nuclear heating.

Analysis of Reactor Power Supply System on Floating Nuclear Power Plants

Chen Qiang, Guo Xiang, Zhu Chenghua

2020, 41(5): 89-93.

Abstract(139)

Abstract:
The safety of the floating nuclear power plant is closely related to the merits of the reactor power supply system. In order to improve the safety factor of floating nuclear power plant, it is necessary to analyze the reactor power supply system. In this paper, the configuration of the reactor power supply system of the floating nuclear power plant is analyzed, and the auxiliary power system and nuclear emergency power system under two schemes are compared. The results show that the optimized scheme 2 is better than scheme 1 in reliability and safety, and scheme 2 is more economical. The optimization scheme proposed in this paper can provide a direction for the design of the reactor power supply system of the subsequent nuclear powered ships and has a strong reference significance.

Research on Monitoring of Reactor Cavity Cooling Status for NPPs

He Peng, Chen Jing, Li Xiaofen, He Zhengxi, Zhu Jialiang, Xu Tao, Li Hongxia

2020, 41(5): 94-98.

Abstract(206)

Abstract:
In order to justify the accident progresses of the cavity and the implementation effect of the cavity injection strategy that is activated under the severe accident condition, the evolution sequence of several cavity physical property parameters of different injection water velocity under severe accident condition is analyzed, and the monitoring measures in the conventional generation 2nd NPP, the improved generation 2nd plus NPP and the HPR1000 NPP are comparatively studied; Based on the optimization of the functional design and algorithm of the temperature measurement instrumentation, water level measurement instrumentation and monitoring system, the monitoring system of the HPR1000 NPP is finally designed. The monitoring system can conduct the accident status monitoring before the failure of the pressure vessel under the severe accident condition, the operation condition monitoring after the activation of the cavity injection strategy, and the molten debris status monitoring after the failure of the pressure vessel, and can satisfy the demand on the cavity status monitoring requirement under the severe accident condition.

Data Reliability Analysis during Containment Leakage Test Based on Statistical Software R

Shen Dongming, Cai Jiantao, He Rui, Huang Xiaoming

2020, 41(5): 99-103.

Abstract(150)

Abstract:
The most important part in the calculation of the containment leakage is to perform the linear regression on time for a series of data measured at different times. The significance test of the regression and residual analysis are the substantial means to evaluate the test results. This paper analyzes the data of the containment test during the commissioning and startup phase of a power plant based on the statistical software R, and explores the regression diagnosis before the leakage calculation by examining the independence, normality and heteroscedasticity of the regression model and the elimination of extreme sample points impact on the reliability of the result. Through the regression diagnosis on the examples, it was found that in the samples which leakage rate is calculated, there may be problems that affect the regression results and then the affect the final results, such as autocorrelation, non-normality and heteroscedasticity. Therefore, the validity of the data shall be evaluated by the regression diagnostic methods, while calculating the leakage rate, and the final results shall be corrected by appropriate methods for samples that fail the test.

Application of Requirement Modeling in Nuclear Requirements Analysis

Zhu Junzhi, Yang Jue, Wan Lei, Cui Jun, Liu Yongkang, Liu Qingsong

2020, 41(5): 104-109.

Abstract(250)

Abstract:
At present, it is increasingly difficult for the nuclear power design products to meet user expectations for there is no effective method to collect and manage the requirement information, and it is unable to perform early the requirement verification and to change the requirements in the research and development process. In view of the above problems, this paper takes the safety injection system as an example and applies the requirement modeling to the requirement analysis process. The collection of requirements is achieved through requirement use case modeling, requirement scenario modeling and requirement logic modeling. the state diagram is implemented to ensure that the top-level design proposal meets user requirements, and the timing diagrams are compared to check for the missing and inconsistent requirements. Therefore, the early validation of requirements is realized by means of requirements modeling, ensuring that the design products meet the user requirements, and at the same time providing a reference for the further application of requirements modeling in nuclear power design.

Simulation Analysis of Boron Concentration Difference between Primary Circuit and Pressurizer of ACPR1000

Jiang Xialan, Li Hui, Qin Zhiguo

2020, 41(5): 110-115.

Abstract(122)

Abstract:
In order to solve the problem of excessively large boron concentration difference between the primary circuit and the pressurizer in the normal dilution and boration process of Improved Chinese Pressurized Water Reactor (ACPR1000), and to prevent the accidental violation of the operating technical specifications of the unit, the cause of the difference in the boron concentration was analyzed theoretically. RELAP5-3D program was used to simulate the reactor neutronics and the thermal hydraulics of the reactor coolant system, and the simulation platform was used to simulate the nuclear island auxiliary and related control systems. The full scope simulator composed of these system models was used to comprehensively analyze the influence of various factors, and quantitatively calculate the difference in the boron concentration under the rapid boration transient. Combined with the operating regulations and experimental data of the unit, the results show that the magnitude of the difference in the boron concentration is greatly affected by the spray and charging flowrate during the transient, and the current operating procedures may lead to that the units exceeds requirements of operation specifications. So that this paper proposes improvements.

Numerical Simulation of Airborne Radioactive Exclusion in Spent Fuel Storage Tank Ventilation Mode

Gui Ting, Xian Chunmei, Fang Zhen, Dong Changqing, An Jing, He Meikui

2020, 41(5): 116-121.

Abstract(134)

Abstract:
In order to protect the workers in the spent fuel storage compartment from internal radiation, the airborne radioactive concentration in the spent fuel storage compartment needs to be controlled. Exhaustion of the airborne radiation is mainly realized through the ventilation system. According to the characteristics of the spent fuel storage tank, four ventilation methods are designed in this paper. Airpak software is used to simulate the four ventilation methods of the spent fuel storage tank. By comparing and analyzing the polluted steam concentration field, flow field and polluted steam trajectory, the effects of four ventilation methods on the airborne radioactive exclusion are studied. The results show that the layered air supply method Ⅱ has a better effect on the airborne radioactive emissions, and under this ventilation mode, the airborne radioactive concentration on the personnel working platform is lower than that in other three ventilation modes.

Research and Risk Analysis of Partial Cooldown Test of EPR Nuclear Power Plant

Zeng Huan, Zhao Xin, Duan Shengzhi

2020, 41(5): 122-126.

Abstract(92)

Abstract:
As the first reactor test of the European Pressurized Water Reactor (EPR) Units, partial cooldown test would cause enormous thermal shocks in the primary and secondary coolant circuits, which are not allowed to happen more than 15 times during the service life of a nuclear power station. In order to lower the risks in the test, studies are furthered on the control logic, theory, principles and the feasibility of the test, and the risks are summarized in 5 aspects. Meanwhile, comprehensive analysis is performed on partial cooldown test by simulations, and then the test method is optimized.The test succeeds at one time, and meets the nuclear safety requirements.

Design and Test Study on a High Temperature Molten Salt Loop

Kong Xiangbo, Wang Naxiu, Lin Liangcheng, Lu Huiju, Fu Yuan, Wang Xiao

2020, 41(5): 127-131.

Abstract(231)

Abstract:
The Thorium Molten Salt Reactor (TMSR) project plans to construct a 2 MWt liquid fuel molten salt reactor. After successful R&D of the proto-type equipment like pump, heat exchanger and freeze valve, construction of a high temperature fluoride loop is designed and constructed to test them. The rated power of heat exchange is designed to be 200 kW, and the flow rate is set as 15 m3/h by the caloric equation. Hence the pipe caliber chose to be DN50. The numerical computation by Fluent revealed that the pressure drop of the loop system is about 155 kPa, and the overdesign of the pump head is set as 20 m accordingly. To avoid thermal stress concentration of the pipeline, both the electric heater and radiator are designed to be fixed on a flexible universal sphere support, except the pump which is anchored on steel support directly. The structural design of the equipment and the loop hase been verified by several tests with an accumulated operating time near 4000 hours. The actual pressure drop of the loop is about 110~120 kPa, which still needs to be accurately measured by differential manometer in future. The analysis of the impurity of the molten salt shows that the contents of Cr and Mo increased by two orders of magnitude after loop operation, which means that an obvious material corrosion occurs. The level of moisture and oxygen, which is the main reason that cause fluoride corrosion, should be controlled considerably lower than the original design level of 100 μL/L.

Research on Condition Monitoring Technology for Nuclear Power Plant Equipment Based on Kernel Principal Component Analysis

Wu Tianhao, Liu Tao, Shi Haining, Zhang Tao, Tang Tang

2020, 41(5): 132-137.

Abstract(333)

Abstract:
In order to solve the limitations of the traditional monitoring methods for nuclear power plants, this paper proposes to introduce Kernel Principal Component Analysis (KPCA) into the online monitoring field of nuclear power plant equipment, and design the monitoring method and online monitoring strategy. In order to verify the effectiveness of the algorithm, it has been applied in the real monitoring case of the motor driven main feed water pump a nuclear power plant in China. The simulation results show that the KPCA algorithm can adapt to the requirements of nuclear power plant equipment monitoring, and can provide earlier warnings of failure than the existing threshold monitoring methods. At the same time, compared with the conventional PCA algorithm, the KPCA algorithm can extract the nonlinear relationship between variables, identify different operating modes of the device, and effectively reduce false alarms.

Research on Lower C Seal Cutting Technology for AP1000 Large Canned Pumps

Yang Zhiye, Li Tao, Wang Binyuan, Li Song, Zhao Mingshen

2020, 41(5): 138-141.

Abstract(135)

Abstract:
The canned motor reactor coolant pumps are used in the Unit 1and Unit 2 in Sanmen Nuclear Power Plant. The lower C seal has to be cut apart by the special way and tool during the pump disassembly. According to the structure of the pump, the functional requirements of the cutting scheme are defined and a cutting toole is designed and developed. The method of finite element analysis (FEA) is used to verify and optimize the structure strength and cooling effect of the tool, to ensure the cutting accuracy and service life. This cutting scheme and the tool are applicable in the narrow and deep space, and the fixed distance cutting can be conducted with high efficiency, high reliability, and without foreign material and air radiation pollution. The scheme and the tool can be used in the nuclear main pumps maintenance in other nuclear power plants, and has certain industrial value.

Model Building Approach for Nuclear Power Operation Procedure Based on Ontology

Xu Yu, Huang Yuanyuan, Xiong Lihong, Leng Shan, Zhu Xiaoliang

2020, 41(5): 142-145.

Abstract(386)

Abstract:
Nuclear power operating procedures are the operational procedures that must be followed for the safe operation of nuclear power plants. Having a lot of texts, the operation procedures are inconvenient for users to browse. However, the current electronic procedures have problems such as lack of simplification, weak content relevance, and less superiority than paper procedures. It is imperative to design a simple and efficient electronic procedure system based Ontology model for nuclear power operation. This paper takes the operating procedures of the chemical and volume systems in a nuclear power plant as an example, uses the ontology model theory, analyzes the text of procedures, develops the knowledge of the procedures and extracts key concepts. Using the ontology development tool to build the system, the electronic procedures and visualization are realized. This construction method is also applicable to the operating procedures of other systems, which not only enables the reuse and sharing of ontology knowledge, but also facilitates querying, understanding, and memorizing procedures. It can help the operators in nuclear power plants to use procedures, thereby improving the efficiency and enhancing the safety and intelligence of nuclear power plants.

Study on Measures to Prevent Pressurizer Overfill During Loss of Normal Feedwater for AP1000 NPP

Ma Baisong, Guo Hongen

2020, 41(5): 146-149.

Abstract(161)

Abstract:
In an accident of loss of feedwater in an AP1000 plant, the pressurizer was filled with water for a series of improper operations, and the safety valves may not be qualified to re-close following multiple cycles of opening, which is not acceptable in Condition Ⅱ events. The paper analyzes the causes for the filling of water in the pressurizer in this event, that is, the instantaneous evaporation of coolant in the loop during the process of improper depressurization of RCS while the PRHR HX is with sufficient cooling capability. At this time, the water level in the pressurizer level cannot be decreased by opening the reactor vessel head vent valves for emergency letdown. Finally, the recommended measure is provided to prevent the filling of water in the pressurizer during loss of normal feedwater for AP1000 NPP. The RCS pressure should always be higher than the saturation pressure corresponding to the temperature of the hot legs to avoid the coolant evaporation.

Research on Impact Factors of Depth Quantification in Eddy Current Test for Thimble Tubes Wear of Nuclear Power Plants

Ma Qiang, Chen Cheng, Li Pingren, Kong Yuying, Ding Boyuan, Zhao Hongqiang, Yang Hongbo

2020, 41(5): 150-154.

Abstract(121)

Abstract:
The result of the eddy current inspection for the thimble tube of a nuclear reactor neutron flux measurement system is the reference for the nuclear power plant to take maintenance actions. Based on the currently-used eddy current inspection method, the effect of the parameter variation of the tube defect on the depth quantification is analyzed. The results show that when the angle of the defect circumferential is less than 210°, it has a greater effect on the eddy current measurement depth quantification and when the defect length is less than 20mm, the effect of the defect depth quantification is obviously. The measurement depth in the eddy current of the crescent-type defect is smaller than the design depth. When the circumferential position of two defects is constant, the overlapped defect has no effect on the measurement, but the overlapped defect will have serious influence for the measurement result if the circumferential position of the two defects changes.

Analysis of Effect of Cone Angle on Performance of Vapor-Anode AMTEC Conical Evaporator

Zhu Lei, Jiang Xinbiao, Li Huaqi, Chen Sen, Tian Xiaoyan, Qiu Suizheng

2020, 41(5): 155-161.

Abstract(122)

Abstract:
To analyze the effect of the cone angle on the performance of the evaporator in vapor-anode Alkali Metal Thermal to Electric Converter (AMTEC), a steady-state two-dimensional thermal hydraulic model for liquid-return wick and conical evaporator was developed. The effects of different cone angles on the surface temperature distribution, void fraction and capillary limit of the evaporator under different operating conditions were studied. The results showed that the evaporation area and vapor pressure along the surface of the conical evaporator were larger than those of the plane evaporator, and the surface vapor pressure of the conical evaporator with 15° cone angle was 35.29 kPa, 1.62 kPa higher than that of the plane evaporator, at cold end temperature of 623 K, hot end temperature of 1123 K and electrical current of 2.5 A. However, the capillary limited domain of the deep cone evaporator was larger than that of the shallow cone evaporator, and it was more likely to experience large area condensation at the center of the evaporator surface. 15° conical evaporator experienced condensation 7%~18% of its surface, at electrical current of 1~2 A. Therefore, the shallow cone evaporator was recommended for the vapor-anode AMTEC.

Effects of Internal Leakage Pathwayon Main Control Room Habitable Dose in Accident Condition of Nuclear Power Plants

Wang Qi, Wang Kai, Wang Jianhua

2020, 41(5): 162-167.

Abstract(185)

Abstract:
In order to fully consider the effects of internal leakage pathway on the radiation safety of the operators in the main control room habitable area during potential accident condition, a radioactive material migration model is established and compartment model is modified considering internal leakage pathways in this paper. The radioactivity in the main control room habitable area and the surrounding non-habitable area is analyzed, and the habitable doses of the operators in the main control room of a nuclear power plant in China are assessed in this paper. At the same time, the sensitivity analysis for the leakage rate of the penetrations is also performed. The results show that: the effective dose and the thyroid equivalent dose of the operators in the habitable area of the main control room calculated by considering the internal leakage pathways method are 8.704 times and 120.749 times of the calculation results without considering the internal leakage pathways. The effective dose increase and the thyroid equivalent dose increase caused by the inhalation pathway are particularly significant. The effective dose is 0.628 mSv and the thyroid equivalent dose is 6.32 mSv which are lower than the limits in HAD 002/01-2019. The sensitivity analysis shows that in the potential accident condition, the effective dose and thyroid equivalent dose of the operators in the main control room habitable area show a clear linear proportional relationship with the penetration rate of penetrations. The research results in this paper provide suggestions for the improvement of the radiation protection design of the main control room and penetrations of nuclear power plants in China.

Research and Application of Calibration Test Method for PTR Liquid Level Gauge in EPR Units

Yuan Meichun, Sun Dongmei, Nan Xiayu

2020, 41(5): 168-172.

Abstract(151)

Abstract:
Aiming at the calibration test of the fuel pool cooling and purification system (PTR) liquid level gauge in EPR units, a test method based on the principle of the connector was proposed, and a special test device was designed. The results show that the new test method can shorten the construction period of the main line by 40 days, reduce the amount of desalination water by more than 3785 tons, and avoid the risk of running water caused by the filling of the whole pool. This method is with low cost, high efficiency, strong practicability and certain promotion value.

Requirements for Monte Carlo Method during Radiation Shielding Optimization Design of Advanced PWR Nuclear Power Plants

Lyu Weifeng, Xiong Jun, Liu Jie, Tang Shaohua

2020, 41(5): 173-177.

Abstract(148)

Abstract:
Based on the analysis of the problems resulting from the tool limitations during the radiation shielding design of M310 nuclear power plant and the requirements presented by the radiation shielding design of HPR1000 nuclear power plant, the requirements for Monte Carlo(MC) method during the optimization design of the radiation shielding for advanced PWR nuclear power plant are presented from four aspects, which are the software interface, input interface, dose rate calculation function and radiation field application expansion. The radiation shielding optimization design practice based on MC method in HPR1000 nuclear power plant shows that, by the development of the software interface, input interfaces and radiation field application expansion, the calculation code based on MC method can play a key role in the radiation shielding optimization design of advanced PWR nuclear power plant, and significantly improve the radiation shielding optimization design of PWR nuclear power plant.

Research of Multi-Objective Optimization Method of Nuclear Reactor Radiation Shielding

Zhang Zehuan, Song Yingming, Lu Chuan, Tang Songqian, Xiao Feng, Lyu Huanwen, Yang Junyun, Mao Jie

2020, 41(5): 178-184.

Abstract(275)

Abstract:
To overcome the disadvantages in the efficiency and applicability of the traditional shielding optimization method based on the Monte Carlo method, in this paper, we studied the reactor radiation shielding optimization method by the non-dominated sorting genetic algorithm (NSGA-Ⅱ) based on the elitist strategy and the mini-batch gradient descent (MBGD), and improved the adaptive mutation rate operator of the genetic algorithm to enhance the global optimization ability of the genetic algorithm. A multi-objective optimization model of the reactor secondary shielding is constructed for comparing the output of normalized neutron transmittance between Monte Carlo method and neural network prediction method, which has verified the accuracy of MBGD. Through the coupling of neural network and NSGA-Ⅱ algorithm, the Pareto front of the radiation shielding design model can be found quickly, which can be applied to the multi-objective optimization engineering design of reactor radiation shielding.

Design Improvement of Reactor Vessel Shielding Component

Zhuang Yaping, Ma Baisong

2020, 41(5): 185-188.

Abstract(122)

Abstract:
During the hot functional test of one NPP, the neutron shielding material was heated and released from the reactor vessel shielding blocks. The structure and layout of the block were redesigned, and B4C was adopted as the neutron shielding material. This paper analyzes the improved design scheme in terms of the heat transfer, the radiation shielding and GSI191. The result indicates that the improved design meet the requirements. During the supplemental hot function test, the temperature of neutron shielding block and module and the radiation dose in the containment were surveyed, and the effectiveness of the new design scheme is further verified.

Overall Design and Verification of ACP100S Floating Nuclear Power Plant

Li Qing, Song Danrong, Zeng Wei, Chen Zhang, Liu Jia, Wang Donghui, Xiao Renjie

2020, 41(5): 189-192.

Abstract(697)

Abstract:
Floating nuclear power plant (FNPP) is a movable nuclear power plant built on the floating platform. FNPP belongs to the category of small reactor based on the classification of electric power. It can be used for power generation, desalination, heating, and can satisfy the special needs of regional power supply, regional heating, offshore oil exploration, remote areas, isolated island, and etc. Based on the analysis of the development trend, the characteristics and the advantages of FNPP, this paper summarizes the ACP100S FNPP solutions of China National Nuclear Corporation, and introduces the design principles, technical features, design parameters, overall technical scheme, test and verification, and demonstration project of ACP100S.

Research on Debris In-Core Cooling and Retention Characteristics

Song Jian, Xiang Qingan, Deng Jian, Yu Hongxing, Du Juan, Bi Jinsheng

2020, 41(5): 193-196.

Abstract(184)

Abstract:
The severe accident analysis model of the small modular reactor ACP100 is built using MELCOR code, and the core heat removed process through the barrel and wall of reactor pressure vessel (RPV) is analyzed by the cavity injection system (CIS). The collapse behavior of the fuel assemblies is estimated by the fuel rod degradation model, and the failure behavior of the lower core plate is estimated by ANSYS program. The results show that the fuel assemblies in the core center melt and collapse to form the core melting pool, while the structure of the fuel assemblies surrounding the core melting pool remains intact, and the core lower plate supports the core melting pool and un-collapsed fuel assemblies all the time, and no creep rupture phenomenon occurs; the core heat can be removed by CIS and the debris in-vessel retention successfully avoids the formation of molten pool in the lower head.

In-pile Performance Simulation and Structure Design of Fully Ceramics Microencapsulated Fuel

Zhou Yi, Liu Shichao, Chen Ping, Li Yuanming, Xin Yong, Liu Zhenhai, Zhang Lin, Gu Mingfei, Zhao Yanli, Le Yunlin

2020, 41(5): 197-200.

Abstract(157)

Abstract:
The thermal mechanical performance of the fully ceramics microencapsulated fuel (FCM) with different non-fuel part size was simulated using two-dimensional characteristic unit. When the fissile loading meet the requirements of the reactor core, the stress condition of SiC matrix and SiC layers were investigated for FCM pellets with different structures. Non-fuel parts and SiC layers suffered relative lower stress by optimizing FCM pellet structure and adjusting distance between different TRISO particles. The stress distribution of matrix, non-fuel part and SiC layer was discussed for the FCM pellets with non-fuel part size from 100 μm to 500 μm. The results indicate that, the maximum hoop stress of the matrix and SiC layer increased with the increasing of non-fuel part size, while the non-fuel parts exhibited crosscurrent. Non-fuel parts and SiC layer possessed lower stress when the non-fuel part was 400 μm. The stress of non-fuel part was about 400 MPa, and the maximum hoop stress of the SiC layers were about 200 MPa. The failure probability was 2.5×10-4. The structure integrity was maintained for the pellets with 400 μm non-fuel part, at the same time the failure probability SiC layer was low. Structural optimization is the basis for the application of FCM pellet.

2020 Vol. 41, No. 5